ML20311A557

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2017 Q1-Q4 ROP Inspection Findings
ML20311A557
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 11/06/2017
From:
Office of Nuclear Reactor Regulation
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Download: ML20311A557 (673)
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[[:#Wiki_filter:1Q/2000 Inspection Findings - Grand Gulf 1 Page 1 of 7 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: N/A Aug 23, 2001

1Q/2000 Inspection Findings - Grand Gulf 1 Page 2 of 7 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing

1Q/2000 Inspection Findings - Grand Gulf 1 Page 3 of 7 The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Barrier Integrity Significance: Sep 27, 2001

1Q/2000 Inspection Findings - Grand Gulf 1 Page 4 of 7 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter

1Q/2000 Inspection Findings - Grand Gulf 1 Page 5 of 7 turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf)

1Q/2000 Inspection Findings - Grand Gulf 1 Page 6 of 7 Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The

1Q/2000 Inspection Findings - Grand Gulf 1 Page 7 of 7 licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Last modified : April 01, 2002

2Q/2000 Inspection Findings - Grand Gulf 1 Page 1 of 7 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: N/A Aug 23, 2001

2Q/2000 Inspection Findings - Grand Gulf 1 Page 2 of 7 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing

2Q/2000 Inspection Findings - Grand Gulf 1 Page 3 of 7 The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Barrier Integrity Significance: Sep 27, 2001

2Q/2000 Inspection Findings - Grand Gulf 1 Page 4 of 7 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in

2Q/2000 Inspection Findings - Grand Gulf 1 Page 5 of 7 the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf)

2Q/2000 Inspection Findings - Grand Gulf 1 Page 6 of 7 Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the

2Q/2000 Inspection Findings - Grand Gulf 1 Page 7 of 7 extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Last modified : April 01, 2002

3Q/2000 Inspection Findings - Grand Gulf 1 Page 1 of 7 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Significance: Dec 29, 2001

3Q/2000 Inspection Findings - Grand Gulf 1 Page 2 of 7 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf)

3Q/2000 Inspection Findings - Grand Gulf 1 Page 3 of 7 Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Barrier Integrity

3Q/2000 Inspection Findings - Grand Gulf 1 Page 4 of 7 Significance: Sep 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey

3Q/2000 Inspection Findings - Grand Gulf 1 Page 5 of 7 of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf)

3Q/2000 Inspection Findings - Grand Gulf 1 Page 6 of 7 Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure

3Q/2000 Inspection Findings - Grand Gulf 1 Page 7 of 7 mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Last modified : March 29, 2002

4Q/2000 Inspection Findings - Grand Gulf 1 Page 1 of 7 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Significance: Dec 29, 2001

4Q/2000 Inspection Findings - Grand Gulf 1 Page 2 of 7 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf)

4Q/2000 Inspection Findings - Grand Gulf 1 Page 3 of 7 Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Barrier Integrity

4Q/2000 Inspection Findings - Grand Gulf 1 Page 4 of 7 Significance: Sep 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey

4Q/2000 Inspection Findings - Grand Gulf 1 Page 5 of 7 of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf)

4Q/2000 Inspection Findings - Grand Gulf 1 Page 6 of 7 Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure

4Q/2000 Inspection Findings - Grand Gulf 1 Page 7 of 7 mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Last modified : March 28, 2002

1Q/2001 Inspection Findings - Grand Gulf 1 Page 1 of 7 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition

1Q/2001 Inspection Findings - Grand Gulf 1 Page 2 of 7 Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation

1Q/2001 Inspection Findings - Grand Gulf 1 Page 3 of 7 Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Barrier Integrity Significance: Sep 27, 2001

1Q/2001 Inspection Findings - Grand Gulf 1 Page 4 of 7 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf) Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events.

1Q/2001 Inspection Findings - Grand Gulf 1 Page 5 of 7 Inspection Report# : 2000004(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf)

1Q/2001 Inspection Findings - Grand Gulf 1 Page 6 of 7 Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color.

1Q/2001 Inspection Findings - Grand Gulf 1 Page 7 of 7 Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Last modified : March 28, 2002

2Q/2001 Inspection Findings - Grand Gulf 1 Page 1 of 7 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001

2Q/2001 Inspection Findings - Grand Gulf 1 Page 2 of 7 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based

2Q/2001 Inspection Findings - Grand Gulf 1 Page 3 of 7 on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Barrier Integrity

2Q/2001 Inspection Findings - Grand Gulf 1 Page 4 of 7 Significance: Sep 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation

2Q/2001 Inspection Findings - Grand Gulf 1 Page 5 of 7 Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf) Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf)

2Q/2001 Inspection Findings - Grand Gulf 1 Page 6 of 7 Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding

2Q/2001 Inspection Findings - Grand Gulf 1 Page 7 of 7 Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Last modified : March 27, 2002

3Q/2001 Inspection Findings - Grand Gulf 1 Page 1 of 7 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for

3Q/2001 Inspection Findings - Grand Gulf 1 Page 2 of 7 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle.

3Q/2001 Inspection Findings - Grand Gulf 1 Page 3 of 7 Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Barrier Integrity Significance: Sep 27, 2001

3Q/2001 Inspection Findings - Grand Gulf 1 Page 4 of 7 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not

3Q/2001 Inspection Findings - Grand Gulf 1 Page 5 of 7 been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf) Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf)

3Q/2001 Inspection Findings - Grand Gulf 1 Page 6 of 7 Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The

3Q/2001 Inspection Findings - Grand Gulf 1 Page 7 of 7 licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Last modified : March 26, 2002

4Q/2001 Inspection Findings - Grand Gulf 1 Page 1 of 6 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material

4Q/2001 Inspection Findings - Grand Gulf 1 Page 2 of 6 permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215).

4Q/2001 Inspection Findings - Grand Gulf 1 Page 3 of 6 Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Barrier Integrity Significance: Sep 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness

4Q/2001 Inspection Findings - Grand Gulf 1 Page 4 of 6 Occupational Radiation Safety Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the

4Q/2001 Inspection Findings - Grand Gulf 1 Page 5 of 6 above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf) Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf) Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding

4Q/2001 Inspection Findings - Grand Gulf 1 Page 6 of 6 Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Last modified : March 01, 2002

1Q/2002 Inspection Findings - Grand Gulf 1 Page 1 of 7 Grand Gulf 1 Initiating Events Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf)

1Q/2002 Inspection Findings - Grand Gulf 1 Page 2 of 7 Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation

1Q/2002 Inspection Findings - Grand Gulf 1 Page 3 of 7 Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Barrier Integrity

1Q/2002 Inspection Findings - Grand Gulf 1 Page 4 of 7 Significance: Sep 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential

1Q/2002 Inspection Findings - Grand Gulf 1 Page 5 of 7 radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf)

1Q/2002 Inspection Findings - Grand Gulf 1 Page 6 of 7 Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000

1Q/2002 Inspection Findings - Grand Gulf 1 Page 7 of 7 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Last modified : July 22, 2002

2Q/2002 Inspection Findings - Grand Gulf 1 Page 1 of 10 Grand Gulf 1 Initiating Events Significance: Jun 29, 2002 Identified By: Self Disclosing Item Type: FIN Finding Failure to perform an adequate technical evaluation for implementation of a modification to the circulating water lubricating water system which nearly initiated a plant transient. Grand Gulf Nuclear Station engineers failed to perform an adequate technical evaluation for Engineering Request 1997-0615-0000 to address the removal of lubricating water flow from upper internals of circulating water Pump A, which had a zero leakage packing installed on the pump shaft. This failure resulted in an unanticipated rise in pump vibrations after commencing maintenance, and placed the plant in a condition where, for 4 minutes, alarm response procedures directed reducing power and securing circulating water Pump A. Condition Report GGNS 2002-0768 was written to document this finding. This finding is more than minor because it was a precursor to a significant event and could have increased the frequency of an initiating event. However, the safety significance was very low (Green) because although an emergency down-power was called for by procedure, increased licensee oversight allowed the operators to restore lubrication and cooling water quickly enough to mitigate the rise in pump vibration eliminating the need for the emergency down-power necessitated by securing the circulating water pump. Inspection Report# : 2002002(pdf) Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Failure to assess and manage the risk associated with main steam line flow transmitter maintenance. Work control center personnel failed to assess and manage the increase in risk for scheduled main steam line flow file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 2 of 10 transmitter maintenance with control rod drive system maintenance already in progress. Concurrent performance of these maintenance jobs would have resulted in the licensee unknowingly placing the plant in a much higher risk condition. Condition Report GGNS 2002-0684 was written to document this inspector finding. This finding is more than minor because it had a potential to create a higher risk condition than was anticipated by the work control center personnel. However, the safety significance was very low (Green) because, upon recognition of the potential for a higher risk condition, work control center personnel canceled the main steam line flow transmitter maintenance and the two maintenance activities were never performed concurrently. Inspection Report# : 2002002(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 3 of 10 Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 4 of 10 Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 5 of 10 Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Barrier Integrity Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Failure to schedule and perform required periodic ultrasonic testing inspections of the residual heat removal system Train A piping elbow following an NRC authorized non-code piping repair. Nondestructive examination personnel failed to perform periodic inspections of a residual heat removal system non-Code repair location, preventing them from determining the rate or extent of future degradation to the elbow location, contrary to the non-Code repair commitment made to the NRC. Condition Report GGNS 2002-0597 was written to document this finding. The finding is more than minor because, following the non-Code piping repair, the nondestructive examination personnel did not have the required ultrasonic test information to diagnose further piping degradation, and may not have taken the appropriate action prior to the development of another residual heat removal file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 6 of 10 system piping through-wall leak. The safety significance of this finding was very low (Green) because, although the elbow wall thickness was not inspected, the subject train was not relied upon for extended operation and the final ultrasonic test results showed no measurable elbow wall thinning. Inspection Report# : 2002002(pdf) Significance: Sep 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674. The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 7 of 10 Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 8 of 10 Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf) Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf) Public Radiation Safety Significance: May 17, 2001 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 9 of 10 Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf) Physical Protection Miscellaneous Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

2Q/2002 Inspection Findings - Grand Gulf 1 Page 10 of 10 Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/03/2003

3Q/2002 Inspection Findings - Grand Gulf 1 Page 1 of 10 Grand Gulf 1 Initiating Events Significance: Jun 29, 2002 Identified By: Self Disclosing Item Type: FIN Finding Failure to perform an adequate technical evaluation for implementation of a modification to the circulating water lubricating water system which nearly initiated a plant transient. Grand Gulf Nuclear Station engineers failed to perform an adequate technical evaluation for Engineering Request 1997-0615-0000 to address the removal of lubricating water flow from upper internals of circulating water Pump A, which had a zero leakage packing installed on the pump shaft. This failure resulted in an unanticipated rise in pump vibrations after commencing maintenance, and placed the plant in a condition where, for 4 minutes, alarm response procedures directed reducing power and securing circulating water Pump A. Condition Report GGNS 2002-0768 was written to document this finding. This finding is more than minor because it was a precursor to a significant event and could have increased the frequency of an initiating event. However, the safety significance was very low (Green) because although an emergency down-power was called for by procedure, increased licensee oversight allowed the operators to restore lubrication and cooling water quickly enough to mitigate the rise in pump vibration eliminating the need for the emergency down-power necessitated by securing the circulating water pump. Inspection Report# : 2002002(pdf) Significance: Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Increased Risk of Loss of Instrument Air System Due to Fire The inspectors identified 33 absorbent pads collecting leaking oil under and around the instrument and service air compressors. No automatic fire detection or suppression equipment was located in the area and only routine inspection of the area was performed by equipment operators once per shift. The absorbent pads were soaked with oil and had the potential to ignite. This increased fire loading without automatic fire detection or suppression capability increased the risk of a loss of instrument air and subsequent reactor scram. Although this issue could be viewed as a precursor to an event, it was determined to have very low risk significance because it did not affect any systems required for safe shutdown of the plant. Inspection Report# : 2000006(pdf) Mitigating Systems Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Failure to assess and manage the risk associated with main steam line flow transmitter maintenance. Work control center personnel failed to assess and manage the increase in risk for scheduled main steam line flow transmitter maintenance with control rod drive system maintenance already in progress. Concurrent performance of these maintenance jobs would have resulted in the licensee unknowingly placing the plant in a much higher risk condition. Condition Report GGNS 2002-0684 was written to document this inspector finding. This finding is more than minor because it had a potential to create a higher risk condition than was anticipated by the work control center

3Q/2002 Inspection Findings - Grand Gulf 1 Page 2 of 10 personnel. However, the safety significance was very low (Green) because, upon recognition of the potential for a higher risk condition, work control center personnel canceled the main steam line flow transmitter maintenance and the two maintenance activities were never performed concurrently. Inspection Report# : 2002002(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: FIN Finding Failure to consider the need to establish goals and monitor the reactor protection system under the maintenance rule program following a functional failure of the end of cycle recirculation pump trip The inspectors determined that the licensee did not recognize that the initiation of the end-of-cycle recirculation pump trip function of the reactor protection system exceeded their Maintenance Rule performance criteria. Therefore, they did not consider the need to establish goals and monitor the system under the Maintenance Rule. This finding is documented in the licensee's corrective action program as Condition Report 2001-1916. This finding had a credible impact on safety because the licensee was unable to trend and establish goals for the system and, therefore, would have limited their ability to determine the effectiveness of the maintenance performed. As a result, the licensee could have experienced future functional failures of the end-of-cycle recirculation pump trip, reducing its reliability. The inspectors determined that the safety significance of this finding was very low (Green). Although the licensee did not consider the need to establish goals and monitor the system, conditions under which the end-of-cycle recirculation pump trip would fail were very limited, all other reactivity control systems remained functional, and the end-of-cycle recirculation pump trip function was not required throughout the remainder of this operating cycle. Inspection Report# : 2001005(pdf) Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to verify adequacy of design of an EOC-RPT modification per 10 CFR 50, Appendix B, Criterion III The inspectors determined that, following a design change modification performed under Engineering Request (ER) 2000-0770, the licensee failed to provide measures for verifying or checking that the end-of-cycle recirculation pump trip function of the reactor protection system was ensured in all cases of turbine control valve fast closure. The modification to the end-of-cycle recirculation pump trip circuitry added margin to the oil pressure set point for the turbine control valve operating oil pressure but made only limited analytical justification relative to short duration turbine control valve fast closures during a short duration load reject. The engineering request did not address all of the inherent timing delays associated with the design of the circuitry installed in the plant. As a result, it remained possible for short duration turbine control valve fast closures to occur without the initiation of an end-of-cycle recirculation pump trip. This was a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and is in the licensee's corrective action program as Condition Report 2001-1371. This violation is more than minor because if a short duration load reject occurred near the end of the operating cycle, the end-of-cycle recirculation pump trip function may not have actuated. The safety significance of this finding was very low (Green) because although the initiation of the end-of-cycle recirculation pump trip function failed, the reactor scrammed with all control rods inserted, the turbine control valves only partially closed, and the turbine bypass valves opened as designed. As a result, the reactor vessel pressure increase was small and had no significant effect on thermal limits. Inspection Report# : 2001005(pdf) Significance: Oct 04, 2002 Identified By: NRC Item Type: NCV NonCited Violation Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system. The licensee failed to establish appropriate instructions for the circumstances when backfilling the reactor core

3Q/2002 Inspection Findings - Grand Gulf 1 Page 3 of 10 isolation cooling high steam flow transmitter. This resulted in technicians improperly backfilling the detector. This caused the detector to isolate steam to the reactor core isolation cooling turbine, rendering the system inoperable. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2002-0947). The finding was of very low safety significance because although the reactor core isolation cooling system was inoperable, all other remaining mitigating systems remained operable and the duration of the system inoperability was short. Inspection Report# : 2002004(pdf) Significance: N/A Aug 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Licensee Identified that failure to follow procedures resulted in Inadequate Combustible Material Controls Technical Specification 5.4.1 requires procedures be established, implemented, and maintained for a Plant Fire Protection Program. Procedure 10-S-03-4, "Fire Prevention: Control of Combustible Material," Revisions 11 and 12 requires the use of combustible material permits to control the storage and use of combustible materials at the Grand Gulf Nuclear Station. Between January 2000 and August 2001, the licensee introduced combustible materials on numerous occasions into areas governed by the applicable procedure without proper use of a combustible material permit. The corrective actions to address this condition were contained in the licensee's condition report CR-GGN-2001-0509. This is being treated as a Non Cited Violation. Inspection Report# : 2001006(pdf) Significance: Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform corrective action in a timely manner. The inspectors determined that the licensee failed to perform adequate corrective actions to replace a control logic relay, for the Division II emergency diesel generator building ventilation system, which was previously identified as being susceptible to failure and required replacement. Failure to replace the subject relay, following previous identified failures, constituted inadequate corrective action and is a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." This violation is noncited in accordance with Section VI.A of NRC's Enforcement Policy and is in the licensee's corrective action program (CR-GGN-2001-1072). This finding also had crosscutting aspects in the area of problem identification and resolution. This finding was of very low safety significance because, although the Division II emergency diesel generator (EDG) outside air fan would not have automatically started in fast speed, the diesel was able to perform it's safety function with the fan in slow speed and room temperature below 120 degrees F and because the operators would still have the opportunity to manually shift the fan to fast speed prior to the room reaching 120 F. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform vendor recommended maintenance. Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering routine preventive maintenance for safety-related equipment. On April 9, 2001, the licensee's root cause analysis determined that their failure to perform vendor recommended routine inspection maintenance of standby liquid control Pump A contributed to the pump's failure and being inoperable as documented in the licensee's corrective action program in Condition Report 2001-0596. This issue is more than minor because standby liquid control Pump A was inoperable for 5 days. This issue was of very low safety significance (Green) because standby liquid control Pump B was available during that time. Inspection Report# : 2001003(pdf) Significance: Jun 30, 2001

3Q/2002 Inspection Findings - Grand Gulf 1 Page 4 of 10 Identified By: NRC Item Type: NCV NonCited Violation Failure to promptly correct a condition that resulted in a degraded heat exchanger surveillance test The licensee's program for evaluating the thermal performance of safety-related heat exchangers identified degraded performance of the risk-significant, safety-related high pressure core spray pump room cooler. The degraded condition of the room cooler occurred between December 12, 1997, and May 22, 2000, for an approximate 30-month period. The condition was eventually determined to be caused by an improper technique for collecting the cooler air flow data, which rendered the cooler capacity software calculation unreliable. The failure to promptly correct the room cooler test deficiencies was a violation of Criterion XVI of Appendix B to 10 CFR Part 50. This violation is being treated as a noncited violation, consistent with Section VI. A of the NRC Enforcement Policy. This violation (50-416/0103-01, Section 1R07) has been entered into the licensee's corrective action program in Condition Report CR-GGN-2001-0591. This finding also had crosscutting aspects in the area of problem identification and resolution. The finding that the high pressure core spray room cooler was degraded was of very low safety significance because all mitigation systems remained operable and barrier integrity was not challenged. Following the inspection, the licensee entered the finding into the corrective action program in Condition Report CR-GGN-2001-0591 Inspection Report# : 2001003(pdf) Significance: Jan 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation Lack of Maintenance Instructions Leads to Overfilling of Standby Diesel Generator Pedestal Bearing The licensee failed to establish adequate instructions to control lube oil replacement and verification of oil level in the pedestal bearing of the Division II standby diesel generator. This resulted in technicians overfilling the bearing oil reservoir which could have resulted in bearing degradation over extended operation. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2001-318). The finding was of very low safety significance because although the pedestal bearing was overfilled, the remaining standby diesel generators and offsite power sources remained operable and the licensee subsequently determined the subject diesel remained capable of performing its design function with the overfilled bearing. Inspection Report# : 2001002(pdf) Significance: Jan 22, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to properly install a kaowool nominal 1-hour fire barrier in the Division II switchgear room (Fire Zone OC215). Operating License Condition 2.C.41 requires Grand Gulf Nuclear Station to meet the requirements of 10 CFR 50, Appendix R, Section III.G.2. Improperly installed Kaowool in the Division II switchgear room did not provide a required nominal 1-hour fire barrier as reported in LER 2000-002, dated January, 22, 2001. The licensee documented this condition in CR 2000-1481. This finding was treated as a noncited violation. This violation was more than minor because if left uncorrected, it would become a more significant safe shutdown safety system availability concern due to the potential loss of both the Division I and II standby service water systems simultaneously. The issue was of very low safety significance because of the relatively low fire ignition frequency, a fire severity factor of 0.1, and the potential for recovery of the affected mitigating systems. Inspection Report# : 2001004(pdf) Significance: Aug 08, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure of the licensee to conduct testing to ensure the continued operability of the high pressure core spray diesel generator ventilation system

3Q/2002 Inspection Findings - Grand Gulf 1 Page 5 of 10 The relay that caused the high pressure core spray diesel generator outside air fan to automatically switch the fan from low to high speed was found to be inoperable since May 2000. A noncited violation of 10 CFR Part 50, Appendix B, Criterion XI was identified for the failure to conduct testing of the high pressure core spray diesel generator ventilation system. This violation is in the licensee's corrective action program as Condition Reports CR-GGN-2000-1115 and 1121. Using the Significance Determination Process, the inspectors determined that the issue was of very low safety significance because the diesel was able to perform it's safety function with the fan in slow speed and because, once the room temperature exceeded 120 F (a temperature measured every shift), operators would have the opportunity to identify that the outside air fan had not automatically shifted and would manually shift the fan to high speed. Inspection Report# : 2000010(pdf) Significance: Jul 01, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to demonstrate control room air conditioning and safeguards switchgear and battery room ventilation systems were effectively controlled through preventive maintenance The inspectors identified numerous control room air conditioning and safeguards switchgear and battery room ventilation system equipment failures that were not adequately evaluated. The licensee did not perform an adequate evaluation of a failure in one train if the other train was available. As a result, the licensee did not determine whether each of the failures resulted from common mode failure causes and was unable to ensure that the systems remained capable of performing their intended function after each failure. By not adequately evaluating the equipment failures, and based on their number, the licensee could not demonstrate that the performance or condition of the systems were being effectively controlled through the performance of appropriate preventive maintenance, as required by the maintenance rule. This was a violation of 10 CFR 50.65(a)(2). This violation (EA-00-167) is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy. This item was entered into the licensee's corrective action program as Condition Report CR-GGN-2000-0809. This issue was determined to have very low risk significance. Both systems are of low risk significance and no Technical Specification limits were exceeded. During the time that each failure occurred, the other train was operable and no actual loss of safety function of safety-related equipment occurred. Inspection Report# : 2000006(pdf) Barrier Integrity Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Failure to schedule and perform required periodic ultrasonic testing inspections of the residual heat removal system Train A piping elbow following an NRC authorized non-code piping repair. Nondestructive examination personnel failed to perform periodic inspections of a residual heat removal system non-Code repair location, preventing them from determining the rate or extent of future degradation to the elbow location, contrary to the non-Code repair commitment made to the NRC. Condition Report GGNS 2002-0597 was written to document this finding. The finding is more than minor because, following the non-Code piping repair, the nondestructive examination personnel did not have the required ultrasonic test information to diagnose further piping degradation, and may not have taken the appropriate action prior to the development of another residual heat removal system piping through-wall leak. The safety significance of this finding was very low (Green) because, although the elbow wall thickness was not inspected, the subject train was not relied upon for extended operation and the final ultrasonic test results showed no measurable elbow wall thinning. Inspection Report# : 2002002(pdf)

3Q/2002 Inspection Findings - Grand Gulf 1 Page 6 of 10 Significance: Sep 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment prior to a planned transient. Technical Specification 5.4.1 requires procedures to be established to provide administrative instructions for equipment controls. On August 27, 2001, the licensee failed to provide adequate prescribed instructions to restore thermal limit monitoring capabilities after deliberately inhibiting the monitoring equipment during a planned transient. The licensee documented this condition in CR 2001-1486. This finding was treated as a noncited violation. The finding had a credible impact on safety because it resulted in the licensee not recognizing operation in a condition exceeding Technical Specification thermal limits in a timely manner. The finding was of very low safety significance because although the finding could have affected the integrity of the fuel cladding, the fuel design limits were not approached and exposure time in this condition was within the limiting condition of operation. Inspection Report# : 2001004(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey on the drywell head On June 6, 2001, the inspector determined that the contamination levels on top of the drywell head were significantly higher than the contamination levels measured around the flange area of the drywell head. The drywell flange area survey results were used as the radiological conditions for the top of the drywell head. These conditions were also used to brief a worker and prescribe protective clothing and respiratory protection prior to assigning work on top of the drywell head. As a result, a worker became contaminated and was assigned an internal dose of 10 millirem. No survey of the top of the drywell head was performed until after the contamination event. The failure to evaluate the concentrations or quantities of radioactive material and the potential radiological hazards on top of the drywell head prior to assigning work in that area is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2001-1088. The safety significance of this violation was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey resulted in an unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a radiological survey of the residual heat removal pump A room An inspector determined that radiation levels in the Residual Heat Removal Pump A room were significantly higher than posted levels due to recent changes in plant operating conditions. The failure to perform a radiological survey is a violation of 10 CFR 20.1501(a). This violation is being treated as a noncited violation consistent with Section VI. A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as CR-GGN-2001-0674.

3Q/2002 Inspection Findings - Grand Gulf 1 Page 7 of 10 The safety significance of the finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The violation was more than minor because the failure to perform a radiological survey has a credible impact on safety and the potential for unplanned or unintended dose. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to wear a proper radiation monitoring device. Technical Specification 5.7.1 requires that an individual permitted to enter a high radiation area be equipped with a radiation monitoring device that continuously integrates dose. On April 21, 2001, the licensee identified that a worker entered a high radiation area with an electronic dosimeter turned off and was, therefore, unable to integrate dose. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0736. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: Apr 27, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform a radiological survey for valve maintenance. 10 CFR 20.1501(a) states that each licensee shall perform surveys that are reasonable to evaluate radiation levels and potential radiological hazards. On April 18, 2001, the licensee identified that during valve maintenance a worker was in a localized area of a larger room that had not been surveyed. This event is described in the licensee's corrective action program, reference Condition Report CR-GGN-2001-0672. This violation is being treated as a noncited violation. The safety significance of this finding was determined to be very low by the Occupational Radiation Safety SDP because there was no overexposure, or substantial potential for an overexposure and the ability to assess dose was not compromised. Inspection Report# : 2001003(pdf) Significance: SL-IV Feb 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform ALARA job reviews for work greater than 1 person-rem On January 31, 2001, the inspector noted two examples where jobs performed during Refueling Outage (RFO) 10 were not reviewed in accordance with station as low as reasonably achievable (ALARA) program procedures. Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.7.2.a. of Procedure 01-S-08-8, "ALARA Program," Revision 16, states in part, that the ALARA Team must review jobs greater than 1 person-rem but less than 5 person-rem. The first example was "Work Inside the Condensers and Hotwells" (RWP 99-09-073), the second was "Emergency Core Cooling System Valve Work" (RWP 99-09-020). Both jobs were originally budgeted for 0.500 person-rem; however, during the work evolution they both exceeded 1 person-rem (1.2 and 2.9 person-rem respectively). The failure of the ALARA Team to review the above jobs that exceeded 1 person-rem was a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation and is in the licensee's corrective action program as Condition Report GGN-2001-0169. The significance of this violation was determined to be more than minor because the failure to perform an appropriate ALARA level review could cause unnecessary additional worker dose, and result in a credible impact on a worker's radiological safety. However, this issue did not affect the cornerstone since there were no overexposures and monitoring devices remained operable. Inspection Report# : 2001002(pdf) Significance: SL-IV Feb 02, 2001

3Q/2002 Inspection Findings - Grand Gulf 1 Page 8 of 10 Identified By: Licensee Item Type: NCV NonCited Violation Failure to perform ALARA re-reviews of jobs previously approved by the ALARA Committee, when the jobs were expected to accumulate greater than 25 percent more than the estimate last approved by the Co Technical Specification 5.4.1 requires procedures for the ALARA program. Section 6.6.6.a.(1) of Procedure 01-S-08-8, "ALARA Program," Revision 16, states, in part, that the ALARA Committee performs re-reviews of jobs previously approved by the Committee, when the jobs are expected to accumulate greater than 25 percent more than the estimate last approved by the Committee. On June 22, 2000, four examples of radiation work permits that exceeded 25 percent more than the last estimate approved by the ALARA Committee were identified, as described in the licensee's corrective action program as CR-GGN-2000-0895. This issue was more than minor but was of very low safety significance because it did not affect a cornerstone. Inspection Report# : 2001002(pdf) Significance: Apr 06, 2000 Identified By: NRC Item Type: NCV NonCited Violation Inadequate survey of items released from the controlled access area The inspectors identified that the licensee failed to adequately survey items released from the controlled access area. Specifically, the licensee failed to evaluate the presence of hard-to-detect radionuclides. The failure to adequately survey items could result in the release of licensed material. This violation of 10 CFR 20.1501(a) is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report CR-GGN-2000-0479. This issue was characterized as a "green" finding based on the occupational radiation safety significance determination process which indicated that the violation had very low risk significance because the violation did not result in public dose greater than 0.005 rem and there were no more than five events. Inspection Report# : 2000004(pdf) Public Radiation Safety Significance: May 17, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to properly classify and manifest a radioactive waste shipment The inspector identified that the licensee excluded measured Pu-241 analysis results from the Reactor Water Cleanup (RWCU) A resin waste stream scaling factors on February 9, 2001. Radioactive Waste Shipment 2001-0203, containing RWCU-A resin, was classified using those scaling factors, manifested, and shipped on February 12, 2001, without determining or documenting the Pu-241 activity. The licensee confirmed that no other shipments during the inspection period were affected. Because the licensee excluded measured Pu-241 activity from their scaling factors, reasonable assurance was not provided that the indirect method of identifying radionuclides in that waste stream was valid. Therefore, the exclusion of Pu-241 in the waste classification and manifest for Radioactive Waste Shipment 2001-0203 was a violation of 10 CFR Part 20, Appendix G. This violation is being treated as a noncited violation consistent with Section VI.A.1 of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Condition Report 2001-0994. The safety significance of this violation was determined to be very low by the Public Radiation Safety SDP because radiation limits were not exceeded and there was no breach of package during transit, certificate of compliance problem, low level burial ground access problem, or failure to make notifications or provide emergency information. The violation was more than minor because there was a credible impact on safety, and it involved an occurrence in the licensee's radioactive material transportation program. Inspection Report# : 2001003(pdf)

3Q/2002 Inspection Findings - Grand Gulf 1 Page 9 of 10 Physical Protection Miscellaneous Significance: N/A Aug 23, 2001 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee adequately identified problems and put them into the corrective action program. The licensee adequately used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Licensee audits and assessments critically assessed the licensee's problem identification and resolution activity and identified needs for improvement in a number of areas including root cause evaluation, timely condition report initiation, and condition report backlogs. During inspection interviews, workers at the site expressed no reservations to input safety issues into the problem identification and resolution program. The licensee implemented corrective actions in a timely manner. The licensee implemented effective corrective actions to prevent recurrence of significant conditions adverse to quality. Inspection Report# : 2001006(pdf) Significance: N/A Sep 13, 2000 Identified By: NRC Item Type: FIN Finding Acceptable Corrective Action Program The licensee was effective at identifying problems and putting them into the corrective action program. The licensee's program was effective at problem identification. The licensee effectively used risk in prioritizing the extent to which individual problems would be evaluated and in establishing schedules for implementation of corrective actions. Corrective actions, when specified, were generally implemented in a timely manner. Licensee audits and assessments were found to be effective. Based on the interviews conducted during this inspection, workers at the site felt free to input safety issues into the problem identification and resolution program. However, the licensee's resolution of two Non-Cited Violations for inadequate corrective actions were narrowly focused or incomplete. The licensee's followup for inadequate corrective actions for main steam isolation valve failures did not identify a potentially generic problem incorporating industry operating experience. The licensee's followup for inadequate corrective actions for repetitive service water check valve failures did not identify that one of the contributing causes was overly narrow searches for similar issues in response to the individual valve failures. Inspection Report# : 2000007(pdf) Significance: N/A Jun 29, 2000 Identified By: NRC Item Type: FIN Finding Incomplete documentation to verify Alert and Notification System performance indicator data The licensee's written documentation of the siren testing data supporting reported performance indicator data was incomplete. The local offsite agencies tested the sirens monthly by conducting a complete cycle of all 43 offsite sirens and orally reported the results to the licensee. The licensee obtained written computer-generated results of monthly siren tests; however, it did not maintain written documentation of alternate methods used by the offsite agencies to determine siren operability, such as local resident verification or growl testing. The licensee entered the issue of incomplete siren test data recording into its corrective action system as Condition Report CR-GGN-2000-0922. This issue was evaluated using the screening process of NRC Inspection Manual Chapter 0609, "Significance Determination Process." By applying the Groups 1, 2, and 3 screening criteria, the inspector determined that the issue did not meet the criteria for entry into the significance determination process because it was not a failure to meet an emergency preparedness planning standard or other regulatory requirement. However, the issue related to the collecting or reporting of performance indicator data. Specifically, the inspector could not verify the performance indicator value from documented test results and determine if a threshold could be exceeded. The inspector concluded that the issue provided substantive information regarding the licensee's ability to conduct an adequate problem identification and

3Q/2002 Inspection Findings - Grand Gulf 1 Page 10 of 10 resolution of siren failures and that the issue had generic implications for other sites that relied on data provided by offsite organizations. By not documenting siren failures in detail, the licensee could not trend failure mechanisms, recurring failures, or the adequacy of corrective actions for previous failures. Therefore, the issue was determined to be a finding of no color. Inspection Report# : 2000006(pdf) Significance: N/A Jun 06, 2000 Identified By: NRC Item Type: FIN Finding Poor Initial Operator License Performance Four of the six initial applicants failed the written examination and overall average scores were low (below passing). This was documented by the licensee in Condition Report CR-GGN-2000-0776. Inspection Report# : 2000301(pdf) Last modified : December 02, 2002

4Q/2002 Inspection Findings - Grand Gulf 1 Page 1 of 4 Grand Gulf 1 Initiating Events Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to prescribe instructions for tightening a reactor recirculation system flange allows unquantifiable torquing of bolts which construct part of the reactor coolant system boundary. A noncited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for failure to establish appropriate instructions for restoration of a reactor recirculation Loop B decontamination flange which resulted in improper torquing of flange bolting and degrading a reactor coolant system pressure boundary. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1988. The noncited violation is greater than minor because it was related to the initiating events cornerstone objective of limiting the likelihood of an initiating event in the form of a loss of coolant from the flanged pressure boundary. The finding was of very low safety significance because although the bolts were improperly torqued and would have been exposed to reactor coolant system pressure, the bolts were replaced by the licensee prior to taking the reactor coolant system to operating pressure due to inspector intervention. Inspection Report# : 2002005(pdf) Significance: Jun 29, 2002 Identified By: Self Disclosing Item Type: FIN Finding Failure to perform an adequate technical evaluation for implementation of a modification to the circulating water lubricating water system which nearly initiated a plant transient. Grand Gulf Nuclear Station engineers failed to perform an adequate technical evaluation for Engineering Request 1997-0615-0000 to address the removal of lubricating water flow from upper internals of circulating water Pump A, which had a zero leakage packing installed on the pump shaft. This failure resulted in an unanticipated rise in pump vibrations after commencing maintenance, and placed the plant in a condition where, for 4 minutes, alarm response procedures directed reducing power and securing circulating water Pump A. Condition Report GGNS 2002-0768 was written to document this finding. This finding is more than minor because it was a precursor to a significant event and could have increased the frequency of an initiating event. However, the safety significance was very low (Green) because although an emergency down-power was called for by procedure, increased licensee oversight allowed the operators to restore lubrication and cooling water quickly enough to mitigate the rise in pump vibration eliminating the need for the emergency down-power necessitated by securing the circulating water pump. Inspection Report# : 2002002(pdf) Mitigating Systems Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions associated with operating the residual heat removal system heat exchanger outlet valve (E12-F003A) beyond its optimum throttling range leads to excessive system vibratio A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for inadequate corrective actions which resulted in operating the residual heat removal system heat exchanger outlet Valve (E12-F003A) beyond its optimum throttling range causing small bore piping failures. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1779. This self-revealing noncited violation is greater than minor because it affected the mitigating systems cornerstone objective of equipment reliability, in that operation of this valve beyond its optimum throttling capability would lead to system small bore piping failures. The finding was of very low safety significance because all other emergency core cooling systems remained available. Inspection Report# : 2002005(pdf) Significance: Dec 27, 2002 Identified By: NRC

4Q/2002 Inspection Findings - Grand Gulf 1 Page 2 of 4 Item Type: NCV NonCited Violation Inadequate design controls associated with adding a permanent pressure locking modification to a residual heat removal system valve resulted in low stress high cycle fatigue whenever the residual heat A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for inadequate design controls which resulted in a pressure locking design modification being completed without provisions for adequate piping supports resulting in a small bore piping failure. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1779. This self-revealing noncited violation is greater than minor because it affected the Mitigating Systems Cornerstone objective of equipment reliability, in that the inadequate design of the pressure locking piping modification allowed cyclic stress to cause a failure of a small bore piping socket weld. The finding was of very low safety significance because all other emergency core cooling systems remained available. Inspection Report# : 2002005(pdf) Significance: Oct 04, 2002 Identified By: NRC Item Type: NCV NonCited Violation Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system. The licensee failed to establish appropriate instructions for the circumstances when backfilling the reactor core isolation cooling high steam flow transmitter. This resulted in technicians improperly backfilling the detector. This caused the detector to isolate steam to the reactor core isolation cooling turbine, rendering the system inoperable. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2002-0947). The finding was of very low safety significance because although the reactor core isolation cooling system was inoperable, all other remaining mitigating systems remained operable and the duration of the system inoperability was short. Inspection Report# : 2002004(pdf) Significance: Sep 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate alternative shutdown procedure A noncited violation of Technical Specification 5.4.1.a was identified for the failure to provide an adequate procedure for a control room fire. Technical Specification 5.4.1.a, requires the licensee to establish procedures for implementation of activities recommended in Regulatory Guide 1.33, which lists procedures for combating a fire in the control room and forced evacuation of the control room. The licensee's Alternative Shutdown Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 25, was inadequate, because it did not instruct operators to verify that a flow diversion pathway was closed, which could render the credited reactor vessel injection source unable to perform its safety function. In the event of a fire in the control room requiring control room evacuation and remote shutdown, this pathway could have diverted coolant to containment spray and away from the reactor vessel through a spuriously opened containment spray valve. Operators would not normally check the valve position on their own and would not have adequate indication from the remote shutdown panel to identify the potential flow diversion path. The licensee entered this finding into their corrective action program as Condition Report CR-GGN-2002-01460. The issue was of greater than minor significance because it impacted the mitigating systems cornerstone and affected the ability of the low pressure coolant injection system to provide adequate core cooling to prevent core damage. Using the Phase 2 Significant Determination Process, this finding was determined to be of very low safety significance, due to the extremely low fire ignition frequency in conjunction with the low probability that fire would cause the spurious opening of the containment spray valve (Section 1R05.3). Inspection Report# : 2002007(pdf) Significance: Sep 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to protect radio repeaters A noncited violation of Grand Gulf Nuclear Station, License Condition 2.C(41), which requires the licensee to implement and maintain the provisions of their NRC-approved fire protection program, was identified. The licensee failed to meet the fire protection program requirement to protect radio repeaters from exposure to fire damage in six fire areas; therefore, in the event of a fire in any one of these fire areas, radio communications necessary to support safe shutdown could be lost. The licensee entered this finding into their corrective action program as Condition Report CR-GGN-2002-1472. The issue was of greater than minor significance because it impacted the mitigating systems cornerstone objective. Specifically, ineffective fire brigade communications can hamper the brigade's ability to fight a fire, thereby, potentially endangering mitigating systems. A Phase 1 Significant Determination Process evaluation determined that the issue has very low safety significance (Green) because the problem only impacts the effectiveness of the fire brigade while other fire protection features, such as fire barriers and physical separation, remain available (Section 1R05.4). Inspection Report# : 2002007(pdf)

4Q/2002 Inspection Findings - Grand Gulf 1 Page 3 of 4 Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Failure to assess and manage the risk associated with main steam line flow transmitter maintenance. Work control center personnel failed to assess and manage the increase in risk for scheduled main steam line flow transmitter maintenance with control rod drive system maintenance already in progress. Concurrent performance of these maintenance jobs would have resulted in the licensee unknowingly placing the plant in a much higher risk condition. Condition Report GGNS 2002-0684 was written to document this inspector finding. This finding is more than minor because it had a potential to create a higher risk condition than was anticipated by the work control center personnel. However, the safety significance was very low (Green) because, upon recognition of the potential for a higher risk condition, work control center personnel canceled the main steam line flow transmitter maintenance and the two maintenance activities were never performed concurrently. Inspection Report# : 2002002(pdf) Barrier Integrity Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Failure to schedule and perform required periodic ultrasonic testing inspections of the residual heat removal system Train A piping elbow following an NRC authorized non-code piping repair. Nondestructive examination personnel failed to perform periodic inspections of a residual heat removal system non-Code repair location, preventing them from determining the rate or extent of future degradation to the elbow location, contrary to the non-Code repair commitment made to the NRC. Condition Report GGNS 2002-0597 was written to document this finding. The finding is more than minor because, following the non-Code piping repair, the nondestructive examination personnel did not have the required ultrasonic test information to diagnose further piping degradation, and may not have taken the appropriate action prior to the development of another residual heat removal system piping through-wall leak. The safety significance of this finding was very low (Green) because, although the elbow wall thickness was not inspected, the subject train was not relied upon for extended operation and the final ultrasonic test results showed no measurable elbow wall thinning. Inspection Report# : 2002002(pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Significance: N/A Dec 06, 2002 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations.

4Q/2002 Inspection Findings - Grand Gulf 1 Page 4 of 4 Inspection Report# : 2002008(pdf) Miscellaneous Last modified : March 25, 2003

1Q/2003 Inspection Findings - Grand Gulf 1 Page 1 of 5 Grand Gulf 1 1Q/2003 Plant Inspection Findings Initiating Events Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to prescribe instructions for tightening a reactor recirculation system flange allows unquantifiable torquing of bolts which construct part of the reactor coolant system boundary. A noncited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for failure to establish appropriate instructions for restoration of a reactor recirculation Loop B decontamination flange which resulted in improper torquing of flange bolting and degrading a reactor coolant system pressure boundary. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1988. The noncited violation is greater than minor because it was related to the initiating events cornerstone objective of limiting the likelihood of an initiating event in the form of a loss of coolant from the flanged pressure boundary. The finding was of very low safety significance because although the bolts were improperly torqued and would have been exposed to reactor coolant system pressure, the bolts were replaced by the licensee prior to taking the reactor coolant system to operating pressure due to inspector intervention. Inspection Report# : 2002005(pdf) Significance: Jun 29, 2002 Identified By: Self Disclosing Item Type: FIN Finding Failure to perform an adequate technical evaluation for implementation of a modification to the circulating water lubricating water system which nearly initiated a plant transient. Grand Gulf Nuclear Station engineers failed to perform an adequate technical evaluation for Engineering Request 1997-0615-0000 to address the removal of lubricating water flow from upper internals of circulating water Pump A, which had a zero leakage packing installed on the pump shaft. This failure resulted in an unanticipated rise in pump vibrations after commencing maintenance, and placed the plant in a condition where, for 4 minutes, alarm response procedures directed reducing power and securing circulating water Pump A. Condition Report GGNS 2002-0768 was written to document this finding. This finding is more than minor because it was a precursor to a significant event and could have increased the frequency of an initiating event. However, the safety significance was very low (Green) because although an emergency down-power was called for by procedure, increased licensee oversight allowed the operators to restore lubrication and cooling water quickly enough to mitigate the rise in pump vibration eliminating the need for the emergency down-power necessitated by securing the circulating water pump. Inspection Report# : 2002002(pdf) Mitigating Systems file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/22/2003

1Q/2003 Inspection Findings - Grand Gulf 1 Page 2 of 5 Significance: Jan 24, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to energize required heat tracing at two fire hose stations located in the emergency diesel generator breeze way during prolonged freezing periods. A noncited violation of Technical Specification 5.4.1.a was identified for failure to provide an adequate administrative procedure for establishing freeze protection measures in the form of heat tracing to fire hose stations located in the emergency diesel generator breezeway. On January 24, 2003, during prolonged freezing temperatures, two fire hose station's heat trace was found unplugged and de-energized. This issue was documented in the licensee's corrective action program as Condition Report CR-GGN-2003-0227. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the mitigating systems cornerstone objective as described in NRC Manual Chapter 0612 involving protection against external factors such as fire. The finding was of very low safety significance because, although the fire hose station's heat trace was not energized, it had not frozen and was restored in a timely manner due to inspector intervention. Inspection Report# : 2002006(pdf) Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions associated with operating the residual heat removal system heat exchanger outlet valve (E12-F003A) beyond its optimum throttling range leads to excessive system vibratio A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for inadequate corrective actions which resulted in operating the residual heat removal system heat exchanger outlet Valve (E12-F003A) beyond its optimum throttling range causing small bore piping failures. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1779. This self-revealing noncited violation is greater than minor because it affected the mitigating systems cornerstone objective of equipment reliability, in that operation of this valve beyond its optimum throttling capability would lead to system small bore piping failures. The finding was of very low safety significance because all other emergency core cooling systems remained available. Inspection Report# : 2002005(pdf) Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate design controls associated with adding a permanent pressure locking modification to a residual heat removal system valve resulted in low stress high cycle fatigue whenever the residual heat A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for inadequate design controls which resulted in a pressure locking design modification being completed without provisions for adequate piping supports resulting in a small bore piping failure. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1779. This self-revealing noncited violation is greater than minor because it affected the Mitigating Systems Cornerstone objective of equipment reliability, in that the inadequate design of the pressure locking piping modification allowed cyclic stress to cause a failure of a small bore piping socket weld. The finding was of very low safety significance because all other emergency core cooling systems remained available. Inspection Report# : 2002005(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/22/2003

1Q/2003 Inspection Findings - Grand Gulf 1 Page 3 of 5 Significance: Oct 04, 2002 Identified By: NRC Item Type: NCV NonCited Violation Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system. The licensee failed to establish appropriate instructions for the circumstances when backfilling the reactor core isolation cooling high steam flow transmitter. This resulted in technicians improperly backfilling the detector. This caused the detector to isolate steam to the reactor core isolation cooling turbine, rendering the system inoperable. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2002-0947). The finding was of very low safety significance because although the reactor core isolation cooling system was inoperable, all other remaining mitigating systems remained operable and the duration of the system inoperability was short. Inspection Report# : 2002004(pdf) Significance: Sep 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate alternative shutdown procedure A noncited violation of Technical Specification 5.4.1.a was identified for the failure to provide an adequate procedure for a control room fire. Technical Specification 5.4.1.a, requires the licensee to establish procedures for implementation of activities recommended in Regulatory Guide 1.33, which lists procedures for combating a fire in the control room and forced evacuation of the control room. The licensee's Alternative Shutdown Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 25, was inadequate, because it did not instruct operators to verify that a flow diversion pathway was closed, which could render the credited reactor vessel injection source unable to perform its safety function. In the event of a fire in the control room requiring control room evacuation and remote shutdown, this pathway could have diverted coolant to containment spray and away from the reactor vessel through a spuriously opened containment spray valve. Operators would not normally check the valve position on their own and would not have adequate indication from the remote shutdown panel to identify the potential flow diversion path. The licensee entered this finding into their corrective action program as Condition Report CR-GGN-2002-01460. The issue was of greater than minor significance because it impacted the mitigating systems cornerstone and affected the ability of the low pressure coolant injection system to provide adequate core cooling to prevent core damage. Using the Phase 2 Significant Determination Process, this finding was determined to be of very low safety significance, due to the extremely low fire ignition frequency in conjunction with the low probability that fire would cause the spurious opening of the containment spray valve (Section 1R05.3). Inspection Report# : 2002007(pdf) Significance: Sep 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to protect radio repeaters A noncited violation of Grand Gulf Nuclear Station, License Condition 2.C(41), which requires the licensee to implement and maintain the provisions of their NRC-approved fire protection program, was identified. The licensee failed to meet the fire protection program requirement to protect radio repeaters from exposure to fire damage in six fire areas; therefore, in the event of a fire in any one of these fire areas, radio communications necessary to support safe shutdown could be lost. The licensee entered this finding into their corrective action program as Condition Report CR-GGN-2002-1472. The issue was of greater than minor significance because it impacted the mitigating systems file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/22/2003

1Q/2003 Inspection Findings - Grand Gulf 1 Page 4 of 5 cornerstone objective. Specifically, ineffective fire brigade communications can hamper the brigade's ability to fight a fire, thereby, potentially endangering mitigating systems. A Phase 1 Significant Determination Process evaluation determined that the issue has very low safety significance (Green) because the problem only impacts the effectiveness of the fire brigade while other fire protection features, such as fire barriers and physical separation, remain available (Section 1R05.4). Inspection Report# : 2002007(pdf) Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Failure to assess and manage the risk associated with main steam line flow transmitter maintenance. Work control center personnel failed to assess and manage the increase in risk for scheduled main steam line flow transmitter maintenance with control rod drive system maintenance already in progress. Concurrent performance of these maintenance jobs would have resulted in the licensee unknowingly placing the plant in a much higher risk condition. Condition Report GGNS 2002-0684 was written to document this inspector finding. This finding is more than minor because it had a potential to create a higher risk condition than was anticipated by the work control center personnel. However, the safety significance was very low (Green) because, upon recognition of the potential for a higher risk condition, work control center personnel canceled the main steam line flow transmitter maintenance and the two maintenance activities were never performed concurrently. Inspection Report# : 2002002(pdf) Barrier Integrity Significance: Jun 29, 2002 Identified By: NRC Item Type: FIN Finding Failure to schedule and perform required periodic ultrasonic testing inspections of the residual heat removal system Train A piping elbow following an NRC authorized non-code piping repair. Nondestructive examination personnel failed to perform periodic inspections of a residual heat removal system non-Code repair location, preventing them from determining the rate or extent of future degradation to the elbow location, contrary to the non-Code repair commitment made to the NRC. Condition Report GGNS 2002-0597 was written to document this finding. The finding is more than minor because, following the non-Code piping repair, the nondestructive examination personnel did not have the required ultrasonic test information to diagnose further piping degradation, and may not have taken the appropriate action prior to the development of another residual heat removal system piping through-wall leak. The safety significance of this finding was very low (Green) because, although the elbow wall thickness was not inspected, the subject train was not relied upon for extended operation and the final ultrasonic test results showed no measurable elbow wall thinning. Inspection Report# : 2002002(pdf) Emergency Preparedness file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/22/2003

1Q/2003 Inspection Findings - Grand Gulf 1 Page 5 of 5 Occupational Radiation Safety Public Radiation Safety Physical Protection Significance: Feb 15, 2003 Identified By: NRC Item Type: NCV NonCited Violation A visitor without unescorted access was found alone without a required escort in a temporary shelter erected in the protected area for inspections of the standby service water system basins. A noncited violation of Section 2.E of the Grand Gulf Nuclear Station (GGNS) facility operating license was identified for failure to comply with Section 6.2, "Access Controls," of the GGNS Security Plan. On February 15, 2003, a GGNS employee, performing access control escort duties, failed to control the access of a visiting contractor who was not authorized by the licensee to enter or remain in the protected area without an escort. This issue was documented in the licensee's corrective action program as Condition Report CR-GGN-2003-0544. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the physical protection cornerstone objective as described in NRC Manual Chapter 0612 involving unescorted visitor access controls. The finding was of very low safety significance because, although the unescorted visitor was found alone, the individual had no intentions of malevolent acts and there had not been two similar findings in the previous four quarters. Inspection Report# : 2002006(pdf) Significance: N/A Dec 06, 2002 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations. Inspection Report# : 2002008(pdf) Miscellaneous Last modified : May 30, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 07/22/2003

2Q/2003 Inspection Findings - Grand Gulf 1 Page 1 of 5 Grand Gulf 1 2Q/2003 Plant Inspection Findings Initiating Events Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to prescribe instructions for tightening a reactor recirculation system flange allows unquantifiable torquing of bolts which construct part of the reactor coolant system boundary. A noncited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for failure to establish appropriate instructions for restoration of a reactor recirculation Loop B decontamination flange which resulted in improper torquing of flange bolting and degrading a reactor coolant system pressure boundary. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1988. The noncited violation is greater than minor because it was related to the initiating events cornerstone objective of limiting the likelihood of an initiating event in the form of a loss of coolant from the flanged pressure boundary. The finding was of very low safety significance because although the bolts were improperly torqued and would have been exposed to reactor coolant system pressure, the bolts were replaced by the licensee prior to taking the reactor coolant system to operating pressure due to inspector intervention. Inspection Report# : 2002005(pdf) Mitigating Systems Significance: Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Apply Adequate Design Control Measures Lead to Increased Agastat Relay Failure Rate The inspectors identified a noncited violation of Criterion III of Appendix B to 10 CFR Part 50 for failure to assure adequate design controls were in place such that Agastat General Purpose relays would be replaced prior to exceeding their design basis life. As a result, 15 out of 17 failed relays in an 18 month period had exceeded their design basis lives; including 4 relays having one or more contacts that would not perform their safety actuation. This finding is greater than minor because, if the condition were left uncorrected it would become a more significant safety concern. Specifically, the affected safety-related systems would have a lower reliability and availability since the failure rate of relays used beyond their service life is significantly higher than those relays that are within their service life. A Significance Determination Process, Phase 3 analysis was performed by the Senior Reactor Analyst in Region IV. It considered the impact of the 4 relays that failed to initiate functions. The 4 relays impacted standby service water to the control room air conditioning system and five containment/drywell isolation valves. The analysis was based on a set of core damage sequences that would initiate from normal operations, but only progress given a loss-of-offsite-power or a loss-of-coolant-accident. The core damage sequence would continue only if the loss of control room air conditioning progressed to a point that control room instrumentation began to fail as a result of high temperatures and operators were required to evacuate the control room. Finally, for core damage to occur, operators would have had to fail to properly shutdown the reactor from the alternate shutdown panel. The analysis indicated that, given this core damage sequence, file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 10/08/2003

2Q/2003 Inspection Findings - Grand Gulf 1 Page 2 of 5 the estimated change in core damage probability was 7.0 x 10-8, and the change in large early release probability was 1.4 x 10-8. The conclusion of this analysis characterized the performance deficiency as an issue of very low safety significance. The licensee implemented an aggressive campaign to replace the affected relays. Inspection Report# : 2003002(pdf) Significance: May 09, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide procedural instructions for restoring the instrument air system following a loss of instrument air. The team identified a noncited violation of Technical Specification 5.4.1 for failure of Grand Gulf Nuclear Station to provide an adequate procedure for restoring the instrument air system following a loss of instrument air. The procedure failed to provide instructions on how to provide seal air and control air to the instrument air compressor from a temporary source. This failure resulted in operation of the unit one instrument air compressor in an abnormal configuration which caused damage to its inlet valve and the licensee's inability to restore instrument air header pressure with that compressor. This issue was documented in the licensee's corrective action program as condition report 2003-1347. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the mitigating systems cornerstone objective as described in NRC Manual Chapter 0612 involving the ability to ensure the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because although the recovery of instrument air was delayed, all mitigating safety system functions remained available. Inspection Report# : 2003007(pdf) Significance: Mar 13, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Determine Cause of Single Relay Contact Failure-to-Transition A noncited violation of Criterion XVI of Appendix B to 10 CFR Part 50 was identified for the failure to adequately identify the cause of relay contact failures-to-transition, a significant condition adverse to quality, and corrective actions to prevent recurrence. This finding is greater than minor because, if the condition were left uncorrected, it would become a more significant safety concern. Specifically, the failure to understand the failure mechanism behind the failure mode mentioned above would impede the licensee's ability to control that failure mechanism and could lead to additional failures of safety-related equipment to actuate when called upon. The finding was determined to be of very low risk significance since no other failures of this type have been experienced since the discovery of the initial five failures Inspection Report# : 2003006(pdf) Significance: Jan 24, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to energize required heat tracing at two fire hose stations located in the emergency diesel generator breeze way during prolonged freezing periods. A noncited violation of Technical Specification 5.4.1.a was identified for failure to provide an adequate administrative procedure for establishing freeze protection measures in the form of heat tracing to fire hose stations located in the emergency diesel generator breezeway. On January 24, 2003, during prolonged freezing temperatures, two fire hose station's heat trace was found unplugged and de-energized. This issue was documented in the licensee's corrective action program as Condition Report CR-GGN-2003-0227. This finding was evaluated using the Significance file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 10/08/2003

2Q/2003 Inspection Findings - Grand Gulf 1 Page 3 of 5 Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the mitigating systems cornerstone objective as described in NRC Manual Chapter 0612 involving protection against external factors such as fire. The finding was of very low safety significance because, although the fire hose station's heat trace was not energized, it had not frozen and was restored in a timely manner due to inspector intervention. Inspection Report# : 2002006(pdf) Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions associated with operating the residual heat removal system heat exchanger outlet valve (E12-F003A) beyond its optimum throttling range leads to excessive system vibratio A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for inadequate corrective actions which resulted in operating the residual heat removal system heat exchanger outlet Valve (E12-F003A) beyond its optimum throttling range causing small bore piping failures. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1779. This self-revealing noncited violation is greater than minor because it affected the mitigating systems cornerstone objective of equipment reliability, in that operation of this valve beyond its optimum throttling capability would lead to system small bore piping failures. The finding was of very low safety significance because all other emergency core cooling systems remained available. Inspection Report# : 2002005(pdf) Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate design controls associated with adding a permanent pressure locking modification to a residual heat removal system valve resulted in low stress high cycle fatigue whenever the residual heat A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for inadequate design controls which resulted in a pressure locking design modification being completed without provisions for adequate piping supports resulting in a small bore piping failure. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1779. This self-revealing noncited violation is greater than minor because it affected the Mitigating Systems Cornerstone objective of equipment reliability, in that the inadequate design of the pressure locking piping modification allowed cyclic stress to cause a failure of a small bore piping socket weld. The finding was of very low safety significance because all other emergency core cooling systems remained available. Inspection Report# : 2002005(pdf) Significance: Oct 04, 2002 Identified By: NRC Item Type: NCV NonCited Violation Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system. The licensee failed to establish appropriate instructions for the circumstances when backfilling the reactor core isolation cooling high steam flow transmitter. This resulted in technicians improperly backfilling the detector. This caused the detector to isolate steam to the reactor core isolation cooling turbine, rendering the system inoperable. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2002-0947). The finding was of very low safety significance because although the reactor core isolation cooling system was inoperable, all other remaining mitigating file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 10/08/2003

2Q/2003 Inspection Findings - Grand Gulf 1 Page 4 of 5 systems remained operable and the duration of the system inoperability was short. Inspection Report# : 2002004(pdf) Significance: Sep 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate alternative shutdown procedure A noncited violation of Technical Specification 5.4.1.a was identified for the failure to provide an adequate procedure for a control room fire. Technical Specification 5.4.1.a, requires the licensee to establish procedures for implementation of activities recommended in Regulatory Guide 1.33, which lists procedures for combating a fire in the control room and forced evacuation of the control room. The licensee's Alternative Shutdown Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 25, was inadequate, because it did not instruct operators to verify that a flow diversion pathway was closed, which could render the credited reactor vessel injection source unable to perform its safety function. In the event of a fire in the control room requiring control room evacuation and remote shutdown, this pathway could have diverted coolant to containment spray and away from the reactor vessel through a spuriously opened containment spray valve. Operators would not normally check the valve position on their own and would not have adequate indication from the remote shutdown panel to identify the potential flow diversion path. The licensee entered this finding into their corrective action program as Condition Report CR-GGN-2002-01460. The issue was of greater than minor significance because it impacted the mitigating systems cornerstone and affected the ability of the low pressure coolant injection system to provide adequate core cooling to prevent core damage. Using the Phase 2 Significant Determination Process, this finding was determined to be of very low safety significance, due to the extremely low fire ignition frequency in conjunction with the low probability that fire would cause the spurious opening of the containment spray valve (Section 1R05.3). Inspection Report# : 2002007(pdf) Significance: Sep 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to protect radio repeaters A noncited violation of Grand Gulf Nuclear Station, License Condition 2.C(41), which requires the licensee to implement and maintain the provisions of their NRC-approved fire protection program, was identified. The licensee failed to meet the fire protection program requirement to protect radio repeaters from exposure to fire damage in six fire areas; therefore, in the event of a fire in any one of these fire areas, radio communications necessary to support safe shutdown could be lost. The licensee entered this finding into their corrective action program as Condition Report CR-GGN-2002-1472. The issue was of greater than minor significance because it impacted the mitigating systems cornerstone objective. Specifically, ineffective fire brigade communications can hamper the brigade's ability to fight a fire, thereby, potentially endangering mitigating systems. A Phase 1 Significant Determination Process evaluation determined that the issue has very low safety significance (Green) because the problem only impacts the effectiveness of the fire brigade while other fire protection features, such as fire barriers and physical separation, remain available (Section 1R05.4). Inspection Report# : 2002007(pdf) Barrier Integrity file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 10/08/2003

2Q/2003 Inspection Findings - Grand Gulf 1 Page 5 of 5 Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Significance: Feb 15, 2003 Identified By: NRC Item Type: NCV NonCited Violation A visitor without unescorted access was found alone without a required escort in a temporary shelter erected in the protected area for inspections of the standby service water system basins. A noncited violation of Section 2.E of the Grand Gulf Nuclear Station (GGNS) facility operating license was identified for failure to comply with Section 6.2, "Access Controls," of the GGNS Security Plan. On February 15, 2003, a GGNS employee, performing access control escort duties, failed to control the access of a visiting contractor who was not authorized by the licensee to enter or remain in the protected area without an escort. This issue was documented in the licensee's corrective action program as Condition Report CR-GGN-2003-0544. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the physical protection cornerstone objective as described in NRC Manual Chapter 0612 involving unescorted visitor access controls. The finding was of very low safety significance because, although the unescorted visitor was found alone, the individual had no intentions of malevolent acts and there had not been two similar findings in the previous four quarters. Inspection Report# : 2002006(pdf) Significance: N/A Dec 06, 2002 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations. Inspection Report# : 2002008(pdf) Miscellaneous Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 10/08/2003

3Q/2003 Inspection Findings - Grand Gulf 1 Page 1 of 5 Grand Gulf 1 3Q/2003 Plant Inspection Findings Initiating Events Significance: Sep 27, 2003 Identified By: NRC Item Type: FIN Finding Failure to identify and resolve a single failure vulnerability contributed to a loss of feedwater event and reactor scram. The inspector identified a self revealing finding because identification and resolution of a single failure vulnerability associated with the condensate system demineralizer isolation valve control circuit was inadequate and contributed to a loss of feedwater event and reactor scram. The licensee documented this finding in their corrective action program as condition report GGNS-CR-2003-300. The finding is greater than minor because it was viewed as a precursor to a significant event and increased the likelihood of an initiating event such as a reactor scram. The finding is of very low safety significance because, although it caused a loss of feedwater event, it did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator; did not contribute to a combination of a reactor trip and loss of mitigation equipment functions; and it did not increase the likelihood of a fire or internal/external flood. Inspection Report# : 2003003(pdf) Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to prescribe instructions for tightening a reactor recirculation system flange allows unquantifiable torquing of bolts which construct part of the reactor coolant system boundary. A noncited violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for failure to establish appropriate instructions for restoration of a reactor recirculation Loop B decontamination flange which resulted in improper torquing of flange bolting and degrading a reactor coolant system pressure boundary. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1988. The noncited violation is greater than minor because it was related to the initiating events cornerstone objective of limiting the likelihood of an initiating event in the form of a loss of coolant from the flanged pressure boundary. The finding was of very low safety significance because although the bolts were improperly torqued and would have been exposed to reactor coolant system pressure, the bolts were replaced by the licensee prior to taking the reactor coolant system to operating pressure due to inspector intervention. Inspection Report# : 2002005(pdf) Mitigating Systems file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 01/12/2004

3Q/2003 Inspection Findings - Grand Gulf 1 Page 2 of 5 Significance: Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Apply Adequate Design Control Measures Lead to Increased Agastat Relay Failure Rate The inspectors identified a noncited violation of Criterion III of Appendix B to 10 CFR Part 50 for failure to assure adequate design controls were in place such that Agastat General Purpose relays would be replaced prior to exceeding their design basis life. As a result, 15 out of 17 failed relays in an 18 month period had exceeded their design basis lives; including 4 relays having one or more contacts that would not perform their safety actuation. This finding is greater than minor because, if the condition were left uncorrected it would become a more significant safety concern. Specifically, the affected safety-related systems would have a lower reliability and availability since the failure rate of relays used beyond their service life is significantly higher than those relays that are within their service life. A Significance Determination Process, Phase 3 analysis was performed by the Senior Reactor Analyst in Region IV. It considered the impact of the 4 relays that failed to initiate functions. The 4 relays impacted standby service water to the control room air conditioning system and five containment/drywell isolation valves. The analysis was based on a set of core damage sequences that would initiate from normal operations, but only progress given a loss-of-offsite-power or a loss-of-coolant-accident. The core damage sequence would continue only if the loss of control room air conditioning progressed to a point that control room instrumentation began to fail as a result of high temperatures and operators were required to evacuate the control room. Finally, for core damage to occur, operators would have had to fail to properly shutdown the reactor from the alternate shutdown panel. The analysis indicated that, given this core damage sequence, the estimated change in core damage probability was 7.0 x 10-8, and the change in large early release probability was 1.4 x 10-8. The conclusion of this analysis characterized the performance deficiency as an issue of very low safety significance. The licensee implemented an aggressive campaign to replace the affected relays. Inspection Report# : 2003002(pdf) Significance: May 09, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide procedural instructions for restoring the instrument air system following a loss of instrument air. The team identified a noncited violation of Technical Specification 5.4.1 for failure of Grand Gulf Nuclear Station to provide an adequate procedure for restoring the instrument air system following a loss of instrument air. The procedure failed to provide instructions on how to provide seal air and control air to the instrument air compressor from a temporary source. This failure resulted in operation of the unit one instrument air compressor in an abnormal configuration which caused damage to its inlet valve and the licensee's inability to restore instrument air header pressure with that compressor. This issue was documented in the licensee's corrective action program as condition report 2003-1347. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the mitigating systems cornerstone objective as described in NRC Manual Chapter 0612 involving the ability to ensure the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because although the recovery of instrument air was delayed, all mitigating safety system functions remained available. Inspection Report# : 2003007(pdf) Significance: Mar 13, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 01/12/2004

3Q/2003 Inspection Findings - Grand Gulf 1 Page 3 of 5 Identified By: NRC Item Type: NCV NonCited Violation Failure to Determine Cause of Single Relay Contact Failure-to-Transition A noncited violation of Criterion XVI of Appendix B to 10 CFR Part 50 was identified for the failure to adequately identify the cause of relay contact failures-to-transition, a significant condition adverse to quality, and corrective actions to prevent recurrence. This finding is greater than minor because, if the condition were left uncorrected, it would become a more significant safety concern. Specifically, the failure to understand the failure mechanism behind the failure mode mentioned above would impede the licensee's ability to control that failure mechanism and could lead to additional failures of safety-related equipment to actuate when called upon. The finding was determined to be of very low risk significance since no other failures of this type have been experienced since the discovery of the initial five failures Inspection Report# : 2003006(pdf) Significance: Jan 24, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to energize required heat tracing at two fire hose stations located in the emergency diesel generator breeze way during prolonged freezing periods. A noncited violation of Technical Specification 5.4.1.a was identified for failure to provide an adequate administrative procedure for establishing freeze protection measures in the form of heat tracing to fire hose stations located in the emergency diesel generator breezeway. On January 24, 2003, during prolonged freezing temperatures, two fire hose station's heat trace was found unplugged and de-energized. This issue was documented in the licensee's corrective action program as Condition Report CR-GGN-2003-0227. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the mitigating systems cornerstone objective as described in NRC Manual Chapter 0612 involving protection against external factors such as fire. The finding was of very low safety significance because, although the fire hose station's heat trace was not energized, it had not frozen and was restored in a timely manner due to inspector intervention. Inspection Report# : 2002006(pdf) Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions associated with operating the residual heat removal system heat exchanger outlet valve (E12-F003A) beyond its optimum throttling range leads to excessive system vibratio A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for inadequate corrective actions which resulted in operating the residual heat removal system heat exchanger outlet Valve (E12-F003A) beyond its optimum throttling range causing small bore piping failures. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1779. This self-revealing noncited violation is greater than minor because it affected the mitigating systems cornerstone objective of equipment reliability, in that operation of this valve beyond its optimum throttling capability would lead to system small bore piping failures. The finding was of very low safety significance because all other emergency core cooling systems remained available. Inspection Report# : 2002005(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 01/12/2004

3Q/2003 Inspection Findings - Grand Gulf 1 Page 4 of 5 Significance: Dec 27, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate design controls associated with adding a permanent pressure locking modification to a residual heat removal system valve resulted in low stress high cycle fatigue whenever the residual heat A noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified for inadequate design controls which resulted in a pressure locking design modification being completed without provisions for adequate piping supports resulting in a small bore piping failure. This issue was documented in the licensee's corrective action program as CR-GGN-2002-1779. This self-revealing noncited violation is greater than minor because it affected the Mitigating Systems Cornerstone objective of equipment reliability, in that the inadequate design of the pressure locking piping modification allowed cyclic stress to cause a failure of a small bore piping socket weld. The finding was of very low safety significance because all other emergency core cooling systems remained available. Inspection Report# : 2002005(pdf) Significance: Oct 04, 2002 Identified By: NRC Item Type: NCV NonCited Violation Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system Performance of maintenance using an inadequate procedure leads to isolation of the reactor core cooling isolation system. The licensee failed to establish appropriate instructions for the circumstances when backfilling the reactor core isolation cooling high steam flow transmitter. This resulted in technicians improperly backfilling the detector. This caused the detector to isolate steam to the reactor core isolation cooling turbine, rendering the system inoperable. This violation of Technical Specification 5.4.1 is noncited in accordance with Section VI.A of the NRC's Enforcement Policy, and is in the licensee's corrective action program (CR-GGN-2002-0947). The finding was of very low safety significance because although the reactor core isolation cooling system was inoperable, all other remaining mitigating systems remained operable and the duration of the system inoperability was short. Inspection Report# : 2002004(pdf) Barrier Integrity Emergency Preparedness file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 01/12/2004

3Q/2003 Inspection Findings - Grand Gulf 1 Page 5 of 5 Occupational Radiation Safety Public Radiation Safety Physical Protection Significance: Feb 15, 2003 Identified By: NRC Item Type: NCV NonCited Violation A visitor without unescorted access was found alone without a required escort in a temporary shelter erected in the protected area for inspections of the standby service water system basins. A noncited violation of Section 2.E of the Grand Gulf Nuclear Station (GGNS) facility operating license was identified for failure to comply with Section 6.2, "Access Controls," of the GGNS Security Plan. On February 15, 2003, a GGNS employee, performing access control escort duties, failed to control the access of a visiting contractor who was not authorized by the licensee to enter or remain in the protected area without an escort. This issue was documented in the licensee's corrective action program as Condition Report CR-GGN-2003-0544. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the physical protection cornerstone objective as described in NRC Manual Chapter 0612 involving unescorted visitor access controls. The finding was of very low safety significance because, although the unescorted visitor was found alone, the individual had no intentions of malevolent acts and there had not been two similar findings in the previous four quarters. Inspection Report# : 2002006(pdf) Significance: N/A Dec 06, 2002 Identified By: NRC Item Type: FIN Finding Verification of Compliance With Interim Compensatory Measures Order On February 25, 2002, the NRC imposed by Order, Interim Compensatory Measures to enhance physical security. The inspectors determined that, overall, the licensee appropriately incorporated the Interim Compensatory Measures into the site protective strategy and access authorization program; developed and implemented relevant procedures; ensured that the emergency plan could be implemented; and established and effectively coordinated interface agreements with offsite organizations. Inspection Report# : 2002008(pdf) Miscellaneous Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 01/12/2004

4Q/2003 Inspection Findings - Grand Gulf 1 Page 1 of 4 Grand Gulf 1 4Q/2003 Plant Inspection Findings Initiating Events Significance: Sep 27, 2003 Identified By: NRC Item Type: FIN Finding Failure to identify and resolve a single failure vulnerability contributed to a loss of feedwater event and reactor scram. The inspector identified a self revealing finding because identification and resolution of a single failure vulnerability associated with the condensate system demineralizer isolation valve control circuit was inadequate and contributed to a loss of feedwater event and reactor scram. The licensee documented this finding in their corrective action program as condition report GGNS-CR-2003-300. The finding is greater than minor because it was viewed as a precursor to a significant event and increased the likelihood of an initiating event such as a reactor scram. The finding is of very low safety significance because, although it caused a loss of feedwater event, it did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator; did not contribute to a combination of a reactor trip and loss of mitigation equipment functions; and it did not increase the likelihood of a fire or internal/external flood. Inspection Report# : 2003003(pdf) Mitigating Systems Significance: Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Apply Adequate Design Control Measures Lead to Increased Agastat Relay Failure Rate The inspectors identified a noncited violation of Criterion III of Appendix B to 10 CFR Part 50 for failure to assure adequate design controls were in place such that Agastat General Purpose relays would be replaced prior to exceeding their design basis life. As a result, 15 out of 17 failed relays in an 18 month period had exceeded their design basis lives; including 4 relays having one or more contacts that would not perform their safety actuation. This finding is greater than minor because, if the condition were left uncorrected it would become a more significant safety concern. Specifically, the affected safety-related systems would have a lower reliability and availability since the failure rate of relays used beyond their service life is significantly higher than those relays that are within their service life. A Significance Determination Process, Phase 3 analysis was performed by the Senior Reactor Analyst in Region IV. It considered the impact of the 4 relays that failed to initiate functions. The 4 relays impacted standby service water to the control room air conditioning system and five containment/drywell isolation valves. The analysis was based on a set of core damage sequences that would initiate from normal operations, but only progress given a loss-of-offsite-power or a loss-of-coolant-accident. The core damage sequence would continue only if the loss of control room air file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 04/22/2004

4Q/2003 Inspection Findings - Grand Gulf 1 Page 2 of 4 conditioning progressed to a point that control room instrumentation began to fail as a result of high temperatures and operators were required to evacuate the control room. Finally, for core damage to occur, operators would have had to fail to properly shutdown the reactor from the alternate shutdown panel. The analysis indicated that, given this core damage sequence, the estimated change in core damage probability was 7.0 x 10-8, and the change in large early release probability was 1.4 x 10-8. The conclusion of this analysis characterized the performance deficiency as an issue of very low safety significance. The licensee implemented an aggressive campaign to replace the affected relays. Inspection Report# : 2003002(pdf) Significance: May 09, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide procedural instructions for restoring the instrument air system following a loss of instrument air. The team identified a noncited violation of Technical Specification 5.4.1 for failure of Grand Gulf Nuclear Station to provide an adequate procedure for restoring the instrument air system following a loss of instrument air. The procedure failed to provide instructions on how to provide seal air and control air to the instrument air compressor from a temporary source. This failure resulted in operation of the unit one instrument air compressor in an abnormal configuration which caused damage to its inlet valve and the licensee's inability to restore instrument air header pressure with that compressor. This issue was documented in the licensee's corrective action program as condition report 2003-1347. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the mitigating systems cornerstone objective as described in NRC Manual Chapter 0612 involving the ability to ensure the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because although the recovery of instrument air was delayed, all mitigating safety system functions remained available. Inspection Report# : 2003007(pdf) Significance: Mar 13, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Determine Cause of Single Relay Contact Failure-to-Transition A noncited violation of Criterion XVI of Appendix B to 10 CFR Part 50 was identified for the failure to adequately identify the cause of relay contact failures-to-transition, a significant condition adverse to quality, and corrective actions to prevent recurrence. This finding is greater than minor because, if the condition were left uncorrected, it would become a more significant safety concern. Specifically, the failure to understand the failure mechanism behind the failure mode mentioned above would impede the licensee's ability to control that failure mechanism and could lead to additional failures of safety-related equipment to actuate when called upon. The finding was determined to be of very low risk significance since no other failures of this type have been experienced since the discovery of the initial five failures Inspection Report# : 2003006(pdf) Significance: Jan 24, 2003 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 04/22/2004

4Q/2003 Inspection Findings - Grand Gulf 1 Page 3 of 4 Failure to energize required heat tracing at two fire hose stations located in the emergency diesel generator breeze way during prolonged freezing periods. A noncited violation of Technical Specification 5.4.1.a was identified for failure to provide an adequate administrative procedure for establishing freeze protection measures in the form of heat tracing to fire hose stations located in the emergency diesel generator breezeway. On January 24, 2003, during prolonged freezing temperatures, two fire hose station's heat trace was found unplugged and de-energized. This issue was documented in the licensee's corrective action program as Condition Report CR-GGN-2003-0227. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the mitigating systems cornerstone objective as described in NRC Manual Chapter 0612 involving protection against external factors such as fire. The finding was of very low safety significance because, although the fire hose station's heat trace was not energized, it had not frozen and was restored in a timely manner due to inspector intervention. Inspection Report# : 2002006(pdf) Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Significance: Feb 15, 2003 Identified By: NRC Item Type: NCV NonCited Violation A visitor without unescorted access was found alone without a required escort in a temporary shelter erected in the protected area for inspections of the standby service water system basins. A noncited violation of Section 2.E of the Grand Gulf Nuclear Station (GGNS) facility operating license was identified for failure to comply with Section 6.2, "Access Controls," of the GGNS Security Plan. On February 15, 2003, a GGNS employee, performing access control escort duties, failed to control the access of a visiting contractor who was not authorized by the licensee to enter or remain in the protected area without an escort. This issue was documented in the licensee's corrective action program as Condition Report CR-GGN-2003-0544. This finding was evaluated using the Significance Determination Process and determined to be of very low safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 04/22/2004

4Q/2003 Inspection Findings - Grand Gulf 1 Page 4 of 4 significance. The finding is greater than minor because it affected the physical protection cornerstone objective as described in NRC Manual Chapter 0612 involving unescorted visitor access controls. The finding was of very low safety significance because, although the unescorted visitor was found alone, the individual had no intentions of malevolent acts and there had not been two similar findings in the previous four quarters. Inspection Report# : 2002006(pdf) Miscellaneous Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\GG1\gg1_pim.html 04/22/2004

1Q/2004 Inspection Findings - Grand Gulf 1 Page 1 of 3 Grand Gulf 1 1Q/2004 Plant Inspection Findings Initiating Events Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Implement Tagging Procedure Resulting in Shutdown of Reactor Water Cleanup System (Section 40A2) The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure of maintenance personnel to comply with a protective tagging procedure while performing work on the reactor water cleanup system. This failure resulted in a leak of reactor coolant requiring an unplanned isolation and shutdown of the reactor water cleanup system. This finding was greater than minor because it affected the human performance attribute of the Initiating Event Cornerstone and affected the cornerstone objective of limiting events that challenge plant stability. The finding was of very low safety significance because it did not increase the likelihood of a loss of coolant accident initiator, did not increase the likelihood of both a reactor trip and unavailability of mitigation equipment, and did not increase the likelihood of a fire or flooding event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Significance: Sep 27, 2003 Identified By: NRC Item Type: FIN Finding Failure to identify and resolve a single failure vulnerability contributed to a loss of feedwater event and reactor scram. The inspector identified a self revealing finding because identification and resolution of a single failure vulnerability associated with the condensate system demineralizer isolation valve control circuit was inadequate and contributed to a loss of feedwater event and reactor scram. The licensee documented this finding in their corrective action program as condition report GGNS-CR-2003-300. The finding is greater than minor because it was viewed as a precursor to a significant event and increased the likelihood of an initiating event such as a reactor scram. The finding is of very low safety significance because, although it caused a loss of feedwater event, it did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator; did not contribute to a combination of a reactor trip and loss of mitigation equipment functions; and it did not increase the likelihood of a fire or internal/external flood. Inspection Report# : 2003003(pdf) Mitigating Systems Significance: Mar 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Store Hydrolazer in Accordance with Design Instructions (Section 1R05) The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for storage of heavy equipment in the containment building in excess of a floor grating capacity contrary to station engineering instructions. This finding was similar to Manual Chapter 0612, Appendix E, Example 4(a). The finding was greater than minor because it adversely affected the containment floor grating yield stress design margin. The licensee's civil engineering staff had to reperform containment structure loading calculations to determine if the subject steel grating could have supported the machine under all loading conditions, including accident conditions. The finding was of very low safety significance because, although the specified grating load rating was exceeded, the new analysis demonstrated that the maximum stresses under accident conditions were below ultimate stress values and the grating would have been capable of supporting the machine under accident conditions. Inspection Report# : 2004002(pdf) 07/14/2004

1Q/2004 Inspection Findings - Grand Gulf 1 Page 2 of 3 Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Maintain Adequate System Operating Instruction to Prevcent Rendering a Required Decay Heat Removal System Inoperable (Section 1R15) The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure to have an adequate electrical bus outage procedure, which resulted in rendering one of two required decay heat removal systems inoperable. This finding was greater than minor because it affected the configuration control attribute of the Mitigating System Cornerstone and affected the cornerstone objective of equipment availability. The finding was of very low safety significance because it did not represent an actual loss of a decay heat removal safety function, did not represent an actual loss of a single train for greater than its allowed Technical Specification outage time, and was not potentially risk significant due to an external initiating event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Corrective Action Results in Through-Wall Corrosion of Ultimate Heat Sink Piping (Section 40A2) The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for failure to correct areas of known localized corrosion prior to the formation of a through-wall leak in the submerged piping of the standby service water system. This finding was greater than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because it did not represent an actual loss of the ultimate heat sink safety function, did not represent an actual loss of a single train for greater than its allowed Technical Specification outage time, and was not potentially risk significant due to an external initiating event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Significance: Jun 28, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Apply Adequate Design Control Measures Lead to Increased Agastat Relay Failure Rate The inspectors identified a noncited violation of Criterion III of Appendix B to 10 CFR Part 50 for failure to assure adequate design controls were in place such that Agastat General Purpose relays would be replaced prior to exceeding their design basis life. As a result, 15 out of 17 failed relays in an 18 month period had exceeded their design basis lives; including 4 relays having one or more contacts that would not perform their safety actuation. This finding is greater than minor because, if the condition were left uncorrected it would become a more significant safety concern. Specifically, the affected safety-related systems would have a lower reliability and availability since the failure rate of relays used beyond their service life is significantly higher than those relays that are within their service life. A Significance Determination Process, Phase 3 analysis was performed by the Senior Reactor Analyst in Region IV. It considered the impact of the 4 relays that failed to initiate functions. The 4 relays impacted standby service water to the control room air conditioning system and five containment/drywell isolation valves. The analysis was based on a set of core damage sequences that would initiate from normal operations, but only progress given a loss-of-offsite-power or a loss-of-coolant-accident. The core damage sequence would continue only if the loss of control room air conditioning progressed to a point that control room instrumentation began to fail as a result of high temperatures and operators were required to evacuate the control room. Finally, for core damage to occur, operators would have had to fail to properly shutdown the reactor from the alternate shutdown panel. The analysis indicated that, given this core damage sequence, the estimated change in core damage probability was 7.0 x 10-8, and the change in large early release probability was 1.4 x 10-8. The conclusion of this analysis characterized the performance deficiency as an issue of very low safety significance. The licensee implemented an aggressive campaign to replace the affected relays. Inspection Report# : 2003002(pdf) Significance: May 09, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to provide procedural instructions for restoring the instrument air system following a loss of instrument air. The team identified a noncited violation of Technical Specification 5.4.1 for failure of Grand Gulf Nuclear Station to provide an adequate 07/14/2004

1Q/2004 Inspection Findings - Grand Gulf 1 Page 3 of 3 procedure for restoring the instrument air system following a loss of instrument air. The procedure failed to provide instructions on how to provide seal air and control air to the instrument air compressor from a temporary source. This failure resulted in operation of the unit one instrument air compressor in an abnormal configuration which caused damage to its inlet valve and the licensee's inability to restore instrument air header pressure with that compressor. This issue was documented in the licensee's corrective action program as condition report 2003-1347. This finding was evaluated using the Significance Determination Process and determined to be of very low safety significance. The finding is greater than minor because it affected the mitigating systems cornerstone objective as described in NRC Manual Chapter 0612 involving the ability to ensure the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because although the recovery of instrument air was delayed, all mitigating safety system functions remained available. Inspection Report# : 2003007(pdf) Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : May 05, 2004 07/14/2004

2Q/2004 Inspection Findings - Grand Gulf 1 Page 1 of 3 Grand Gulf 1 2Q/2004 Plant Inspection Findings Initiating Events Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failute to Implement Surveillance Procedure Resulting in the Inadvertent Initiation of HPCS System The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure of maintenance technicians to comply with a surveillance procedure for performing maintenance on the reactor vessel water level control system. This failure resulted in the high pressure core spray system inadvertently initiating and injecting into the reactor vessel. This finding is greater than minor because it affected the human performance attribute (human error) of the Initiating Events Cornerstone and affected the cornerstone objective of limiting events that challenge plant stability. The finding was of very low safety significance because it did not contribute to the likelihood of a primary or secondary loss of coolant accident initiator; did not contribute to both the likelihood of a reactor trip and the likelihood of the mitigation equipment or functions being unavailable; nor did it increase the likelihood of a fire or internal/external flooding. Inspection Report# : 2004003(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Implement Tagging Procedure Resulting in Shutdown of Reactor Water Cleanup System (Section 40A2) The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure of maintenance personnel to comply with a protective tagging procedure while performing work on the reactor water cleanup system. This failure resulted in a leak of reactor coolant requiring an unplanned isolation and shutdown of the reactor water cleanup system. This finding was greater than minor because it affected the human performance attribute of the Initiating Event Cornerstone and affected the cornerstone objective of limiting events that challenge plant stability. The finding was of very low safety significance because it did not increase the likelihood of a loss of coolant accident initiator, did not increase the likelihood of both a reactor trip and unavailability of mitigation equipment, and did not increase the likelihood of a fire or flooding event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Significance: Sep 27, 2003 Identified By: NRC Item Type: FIN Finding Failure to identify and resolve a single failure vulnerability contributed to a loss of feedwater event and reactor scram. The inspector identified a self revealing finding because identification and resolution of a single failure vulnerability associated with the condensate system demineralizer isolation valve control circuit was inadequate and contributed to a loss of feedwater event and reactor scram. The licensee documented this finding in their corrective action program as condition report GGNS-CR-2003-300. The finding is greater than minor because it was viewed as a precursor to a significant event and increased the likelihood of an initiating event such as a reactor scram. The finding is of very low safety significance because, although it caused a loss of feedwater event, it did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator; did not contribute to a combination of a reactor trip and loss of mitigation equipment functions; and it did not increase the likelihood of a fire or internal/external flood. Inspection Report# : 2003003(pdf) Mitigating Systems Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation

2Q/2004 Inspection Findings - Grand Gulf 1 Page 2 of 3 Improper Valve Lineup Results in Isolation of RHR Pump Minimum Flow Line A self-revealing Green noncited violation of Technical Specification 5.4.1.a involved the failure of operators to comply with a valve lineup procedure prior to restoring the residual heat removal system to operation. This failure resulted in the isolation of the minimum flow line for the Train B residual heat removal pump, rendering one low pressure emergency core cooling system inoperable for 14 days, which violated the requirements of Technical Specification 3.5.1 prohibiting power operation with one low pressure emergency core cooling system out of service for greater than 7 days. This finding is greater than minor because it affected the configuration control and human performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. Using the Inspection Manual Chapter 0609 Significance Determination Process Phase 1 screening worksheet, this performance deficiency required a Phase 2 evaluation since it resulted in the actual loss of a single train for longer than its Technical Specification Allowed Outage Time. The Phase 2 and Phase 3 evaluations determined this finding to result in a core damage frequency change of less than 1.0E-6 and a change in Large Early Release Fraction of less than 1.0E-7. Therefore, the finding was considered to be of very low safety significance. Inspection Report# : 2004003(pdf) Significance: Mar 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Store Hydrolazer in Accordance with Design Instructions (Section 1R05) The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for storage of heavy equipment in the containment building in excess of a floor grating capacity contrary to station engineering instructions. This finding was similar to Manual Chapter 0612, Appendix E, Example 4(a). The finding was greater than minor because it adversely affected the containment floor grating yield stress design margin. The licensee's civil engineering staff had to reperform containment structure loading calculations to determine if the subject steel grating could have supported the machine under all loading conditions, including accident conditions. The finding was of very low safety significance because, although the specified grating load rating was exceeded, the new analysis demonstrated that the maximum stresses under accident conditions were below ultimate stress values and the grating would have been capable of supporting the machine under accident conditions. Inspection Report# : 2004002(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Maintain Adequate System Operating Instruction to Prevcent Rendering a Required Decay Heat Removal System Inoperable (Section 1R15) The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure to have an adequate electrical bus outage procedure, which resulted in rendering one of two required decay heat removal systems inoperable. This finding was greater than minor because it affected the configuration control attribute of the Mitigating System Cornerstone and affected the cornerstone objective of equipment availability. The finding was of very low safety significance because it did not represent an actual loss of a decay heat removal safety function, did not represent an actual loss of a single train for greater than its allowed Technical Specification outage time, and was not potentially risk significant due to an external initiating event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Corrective Action Results in Through-Wall Corrosion of Ultimate Heat Sink Piping (Section 40A2) The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for failure to correct areas of known localized corrosion prior to the formation of a through-wall leak in the submerged piping of the standby service water system. This finding was greater than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because it did not represent an actual loss of the ultimate heat sink safety function, did not represent an actual loss of a single train for greater than its allowed Technical Specification outage time, and was not potentially risk significant due to an external initiating event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Barrier Integrity

2Q/2004 Inspection Findings - Grand Gulf 1 Page 3 of 3 Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Implement Reactor Startup Procedure Resulting in the Inadvertent Misalignment of the Control Rod Pattern The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a as a result of reactor operators failing to comply with an operating procedure used to establish a required rod pattern configuration during a reactor startup. This failure resulted in the reactor operators inadvertently withdrawing a control rod out of sequence. This finding is greater than minor because it involved the configuration control attribute (reactivity control) of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because it only affected the fuel barrier and not the reactor coolant system barrier. Inspection Report# : 2004003(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Jun 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Follow a Radiation Work Permit Requirement A self-revealing noncited violation of Technical Specification 5.4.1.a was evaluated for a worker who failed to follow a radiation work permit requirement. On March 15, 2004, a worker alarmed the personnel contamination monitors upon exiting the Radiologically Controlled Area because the individual had become contaminated. A follow-up survey of the work area identified contamination levels of up to 180,000 disintegrations per minute per 100 cm2 inside a drain pipe and 500,000 disintegrations per minute per 100 cm2 inside the valve housing. The licensee determined that the worker did not follow the radiation work permit requirement to contact Radiation Protection for approval before commencing cutting activities. This finding is greater than minor because it is associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding was of very low safety significance because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. Inspection Report# : 2004003(pdf) Public Radiation Safety Physical Protection Physical Protection information not publicly available. Miscellaneous Last modified : September 08, 2004

3Q/2004 Inspection Findings - Grand Gulf 1 Page 1 of 3 Grand Gulf 1 3Q/2004 Plant Inspection Findings Initiating Events Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failute to Implement Surveillance Procedure Resulting in the Inadvertent Initiation of HPCS System The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure of maintenance technicians to comply with a surveillance procedure for performing maintenance on the reactor vessel water level control system. This failure resulted in the high pressure core spray system inadvertently initiating and injecting into the reactor vessel. This finding is greater than minor because it affected the human performance attribute (human error) of the Initiating Events Cornerstone and affected the cornerstone objective of limiting events that challenge plant stability. The finding was of very low safety significance because it did not contribute to the likelihood of a primary or secondary loss of coolant accident initiator; did not contribute to both the likelihood of a reactor trip and the likelihood of the mitigation equipment or functions being unavailable; nor did it increase the likelihood of a fire or internal/external flooding. Inspection Report# : 2004003(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Implement Tagging Procedure Resulting in Shutdown of Reactor Water Cleanup System (Section 40A2) The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure of maintenance personnel to comply with a protective tagging procedure while performing work on the reactor water cleanup system. This failure resulted in a leak of reactor coolant requiring an unplanned isolation and shutdown of the reactor water cleanup system. This finding was greater than minor because it affected the human performance attribute of the Initiating Event Cornerstone and affected the cornerstone objective of limiting events that challenge plant stability. The finding was of very low safety significance because it did not increase the likelihood of a loss of coolant accident initiator, did not increase the likelihood of both a reactor trip and unavailability of mitigation equipment, and did not increase the likelihood of a fire or flooding event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Mitigating Systems Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Improper Valve Lineup Results in Isolation of RHR Pump Minimum Flow Line A self-revealing Green noncited violation of Technical Specification 5.4.1.a involved the failure of operators to comply with a valve lineup procedure prior to restoring the residual heat removal system to operation. This failure resulted in the isolation of the minimum flow line for the Train B residual heat removal pump, rendering one low pressure emergency core cooling system inoperable for 14 days, which violated the requirements of Technical Specification 3.5.1 prohibiting power operation with one low pressure emergency core cooling system out of service for greater than 7 days. This finding is greater than minor because it affected the configuration control and human performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. Using the Inspection Manual Chapter 0609 Significance Determination Process Phase 1 screening worksheet, this performance deficiency required a Phase 2 evaluation since it resulted in the actual loss of a single train for longer than its Technical Specification Allowed Outage Time. The Phase 2 and Phase 3 evaluations determined this finding to result in a core damage frequency change of less than 1.0E-6 and a change in Large Early Release Fraction of less than 1.0E-7. Therefore, the finding was considered to be of very low safety significance.

3Q/2004 Inspection Findings - Grand Gulf 1 Page 2 of 3 Inspection Report# : 2004003(pdf) Significance: Mar 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Store Hydrolazer in Accordance with Design Instructions (Section 1R05) The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for storage of heavy equipment in the containment building in excess of a floor grating capacity contrary to station engineering instructions. This finding was similar to Manual Chapter 0612, Appendix E, Example 4(a). The finding was greater than minor because it adversely affected the containment floor grating yield stress design margin. The licensee's civil engineering staff had to reperform containment structure loading calculations to determine if the subject steel grating could have supported the machine under all loading conditions, including accident conditions. The finding was of very low safety significance because, although the specified grating load rating was exceeded, the new analysis demonstrated that the maximum stresses under accident conditions were below ultimate stress values and the grating would have been capable of supporting the machine under accident conditions. Inspection Report# : 2004002(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Maintain Adequate System Operating Instruction to Prevcent Rendering a Required Decay Heat Removal System Inoperable (Section 1R15) The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure to have an adequate electrical bus outage procedure, which resulted in rendering one of two required decay heat removal systems inoperable. This finding was greater than minor because it affected the configuration control attribute of the Mitigating System Cornerstone and affected the cornerstone objective of equipment availability. The finding was of very low safety significance because it did not represent an actual loss of a decay heat removal safety function, did not represent an actual loss of a single train for greater than its allowed Technical Specification outage time, and was not potentially risk significant due to an external initiating event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Corrective Action Results in Through-Wall Corrosion of Ultimate Heat Sink Piping (Section 40A2) The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for failure to correct areas of known localized corrosion prior to the formation of a through-wall leak in the submerged piping of the standby service water system. This finding was greater than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because it did not represent an actual loss of the ultimate heat sink safety function, did not represent an actual loss of a single train for greater than its allowed Technical Specification outage time, and was not potentially risk significant due to an external initiating event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Barrier Integrity Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Implement Reactor Startup Procedure Resulting in the Inadvertent Misalignment of the Control Rod Pattern The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a as a result of reactor operators failing to comply with an operating procedure used to establish a required rod pattern configuration during a reactor startup. This failure resulted in the reactor

3Q/2004 Inspection Findings - Grand Gulf 1 Page 3 of 3 operators inadvertently withdrawing a control rod out of sequence. This finding is greater than minor because it involved the configuration control attribute (reactivity control) of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because it only affected the fuel barrier and not the reactor coolant system barrier. Inspection Report# : 2004003(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Jun 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Follow a Radiation Work Permit Requirement A self-revealing noncited violation of Technical Specification 5.4.1.a was evaluated for a worker who failed to follow a radiation work permit requirement. On March 15, 2004, a worker alarmed the personnel contamination monitors upon exiting the Radiologically Controlled Area because the individual had become contaminated. A follow-up survey of the work area identified contamination levels of up to 180,000 disintegrations per minute per 100 cm2 inside a drain pipe and 500,000 disintegrations per minute per 100 cm2 inside the valve housing. The licensee determined that the worker did not follow the radiation work permit requirement to contact Radiation Protection for approval before commencing cutting activities. This finding is greater than minor because it is associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding was of very low safety significance because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. Inspection Report# : 2004003(pdf) Public Radiation Safety Physical Protection Physical Protection information not publicly available. Miscellaneous Last modified : December 29, 2004

4Q/2004 Inspection Findings - Grand Gulf 1 Page 1 of 4 Grand Gulf 1 4Q/2004 Plant Inspection Findings Initiating Events Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failute to Implement Surveillance Procedure Resulting in the Inadvertent Initiation of HPCS System The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure of maintenance technicians to comply with a surveillance procedure for performing maintenance on the reactor vessel water level control system. This failure resulted in the high pressure core spray system inadvertently initiating and injecting into the reactor vessel. This finding is greater than minor because it affected the human performance attribute (human error) of the Initiating Events Cornerstone and affected the cornerstone objective of limiting events that challenge plant stability. The finding was of very low safety significance because it did not contribute to the likelihood of a primary or secondary loss of coolant accident initiator; did not contribute to both the likelihood of a reactor trip and the likelihood of the mitigation equipment or functions being unavailable; nor did it increase the likelihood of a fire or internal/external flooding. Inspection Report# : 2004003(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Implement Tagging Procedure Resulting in Shutdown of Reactor Water Cleanup System (Section 40A2) The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure of maintenance personnel to comply with a protective tagging procedure while performing work on the reactor water cleanup system. This failure resulted in a leak of reactor coolant requiring an unplanned isolation and shutdown of the reactor water cleanup system. This finding was greater than minor because it affected the human performance attribute of the Initiating Event Cornerstone and affected the cornerstone objective of limiting events that challenge plant stability. The finding was of very low safety significance because it did not increase the likelihood of a loss of coolant accident initiator, did not increase the likelihood of both a reactor trip and unavailability of mitigation equipment, and did not increase the likelihood of a fire or flooding event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Mitigating Systems Significance: Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to retain safety related records for installation and testing of diesel generator starting air storage tank relief valves The inspectors identified a noncited violation of 10 CFR Part 50.71, "Maintenance of Records, Making of Reports," for failure of the licensee to retain safety related records relating to the periodic testing of the high pressure core spray emergency diesel generator starting air storage tank relief valves. This finding is more than minor because it is analogous to example 1.b of Appendix E of IMC 0612, "Power Reactor Inspection Reports," in that the safety related records were irretrievably lost. Using the Significance Determination Process Phase 1 worksheet, the inspectors determined the finding affected the mitigating systems cornerstone and was of very low safety significance because it did not represent an actual loss of system function. Inspection Report# : 2004005(pdf) Significance: Jun 30, 2004 Identified By: Self Disclosing

4Q/2004 Inspection Findings - Grand Gulf 1 Page 2 of 4 Item Type: NCV NonCited Violation Improper Valve Lineup Results in Isolation of RHR Pump Minimum Flow Line A self-revealing Green noncited violation of Technical Specification 5.4.1.a involved the failure of operators to comply with a valve lineup procedure prior to restoring the residual heat removal system to operation. This failure resulted in the isolation of the minimum flow line for the Train B residual heat removal pump, rendering one low pressure emergency core cooling system inoperable for 14 days, which violated the requirements of Technical Specification 3.5.1 prohibiting power operation with one low pressure emergency core cooling system out of service for greater than 7 days. This finding is greater than minor because it affected the configuration control and human performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. Using the Inspection Manual Chapter 0609 Significance Determination Process Phase 1 screening worksheet, this performance deficiency required a Phase 2 evaluation since it resulted in the actual loss of a single train for longer than its Technical Specification Allowed Outage Time. The Phase 2 and Phase 3 evaluations determined this finding to result in a core damage frequency change of less than 1.0E-6 and a change in Large Early Release Fraction of less than 1.0E-7. Therefore, the finding was considered to be of very low safety significance. Inspection Report# : 2004003(pdf) Significance: Mar 27, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Store Hydrolazer in Accordance with Design Instructions (Section 1R05) The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for storage of heavy equipment in the containment building in excess of a floor grating capacity contrary to station engineering instructions. This finding was similar to Manual Chapter 0612, Appendix E, Example 4(a). The finding was greater than minor because it adversely affected the containment floor grating yield stress design margin. The licensee's civil engineering staff had to reperform containment structure loading calculations to determine if the subject steel grating could have supported the machine under all loading conditions, including accident conditions. The finding was of very low safety significance because, although the specified grating load rating was exceeded, the new analysis demonstrated that the maximum stresses under accident conditions were below ultimate stress values and the grating would have been capable of supporting the machine under accident conditions. Inspection Report# : 2004002(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Maintain Adequate System Operating Instruction to Prevcent Rendering a Required Decay Heat Removal System Inoperable (Section 1R15) The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure to have an adequate electrical bus outage procedure, which resulted in rendering one of two required decay heat removal systems inoperable. This finding was greater than minor because it affected the configuration control attribute of the Mitigating System Cornerstone and affected the cornerstone objective of equipment availability. The finding was of very low safety significance because it did not represent an actual loss of a decay heat removal safety function, did not represent an actual loss of a single train for greater than its allowed Technical Specification outage time, and was not potentially risk significant due to an external initiating event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf) Significance: Mar 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Inadequate Corrective Action Results in Through-Wall Corrosion of Ultimate Heat Sink Piping (Section 40A2) The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for failure to correct areas of known localized corrosion prior to the formation of a through-wall leak in the submerged piping of the standby service water system. This finding was greater than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The finding was of very low safety significance because it did not represent an actual loss of the ultimate heat sink safety function, did not represent an actual loss of a single train for greater than its allowed Technical Specification outage time, and was not potentially risk significant due to an external initiating event as described in the significance determination process Phase 1 screening worksheet. Inspection Report# : 2004002(pdf)

4Q/2004 Inspection Findings - Grand Gulf 1 Page 3 of 4 Barrier Integrity Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Implement Reactor Startup Procedure Resulting in the Inadvertent Misalignment of the Control Rod Pattern The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a as a result of reactor operators failing to comply with an operating procedure used to establish a required rod pattern configuration during a reactor startup. This failure resulted in the reactor operators inadvertently withdrawing a control rod out of sequence. This finding is greater than minor because it involved the configuration control attribute (reactivity control) of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because it only affected the fuel barrier and not the reactor coolant system barrier. Inspection Report# : 2004003(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to follow a source calibration procedure resulting in a worker receiving an unplanned, unintended dose. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1(a) for a worker who failed to follow a source calibration procedure and removed a lead attenuator while the radioactive source was in the up (exposed) position. As a result, the worker unintentionally exposed himself to a dose rate of approximately 330 millirem/hour, received an unplanned dose of one millirem and had the potential to receive additional unnecessary dose. This finding is greater than minor since it involves a worker's unplanned, unintended dose resulting from actions contrary to licensee procedures, which is associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and directly affects the cornerstone objective to ensure adequate protection of the worker's health and safety from exposure to radiation. The inspectors evaluated the finding using the Occupational Radiation Safety Significance Determination Process and determined it was of very low safety significance because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. Inspection Report# : 2004005(pdf) Significance: Jun 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Follow a Radiation Work Permit Requirement A self-revealing noncited violation of Technical Specification 5.4.1.a was evaluated for a worker who failed to follow a radiation work permit requirement. On March 15, 2004, a worker alarmed the personnel contamination monitors upon exiting the Radiologically Controlled Area because the individual had become contaminated. A follow-up survey of the work area identified contamination levels of up to 180,000 disintegrations per minute per 100 cm2 inside a drain pipe and 500,000 disintegrations per minute per 100 cm2 inside the valve housing. The licensee determined that the worker did not follow the radiation work permit requirement to contact Radiation Protection for approval before commencing cutting activities. This finding is greater than minor because it is associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding was of very low safety significance because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose.

4Q/2004 Inspection Findings - Grand Gulf 1 Page 4 of 4 Inspection Report# : 2004003(pdf) Public Radiation Safety Physical Protection Physical Protection information not publicly available. Miscellaneous Last modified : March 09, 2005

1Q/2005 Inspection Findings - Grand Gulf 1 Page 1 of 4 Grand Gulf 1 1Q/2005 Plant Inspection Findings Initiating Events Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failute to Implement Surveillance Procedure Resulting in the Inadvertent Initiation of HPCS System The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for failure of maintenance technicians to comply with a surveillance procedure for performing maintenance on the reactor vessel water level control system. This failure resulted in the high pressure core spray system inadvertently initiating and injecting into the reactor vessel. This finding is greater than minor because it affected the human performance attribute (human error) of the Initiating Events Cornerstone and affected the cornerstone objective of limiting events that challenge plant stability. The finding was of very low safety significance because it did not contribute to the likelihood of a primary or secondary loss of coolant accident initiator; did not contribute to both the likelihood of a reactor trip and the likelihood of the mitigation equipment or functions being unavailable; nor did it increase the likelihood of a fire or internal/external flooding. Inspection Report# : 2004003(pdf) Mitigating Systems Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Operability Dertmination Procedure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for a failure to follow procedures that resulted in an inadequate operability determination. Specifically, operators failed to adequately implement the provisions of their operability determination to evaluate a degraded condition in the control room air conditioning system. This finding was greater than minor since it is associated with the equipment performance attribute of the mitigating systems cornerstone and directly affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process," the finding was of very low safety significance since: (1) it did not represent an actual loss of system safety function for the control room air conditioning system, (2) it did result in a loss of function for a single train of Technical Specification equipment, but for less than the Technical Specification allowed outage time, and (3) it did not represent a loss of function of non-technical specification risk significant equipment or screen as potentially risk significant due to a seismic, flooding or severe weather event. This finding has crosscutting aspects associated with human performance in that the control room operators failed to implement the operability determination procedure. Inspection Report# : 2005002(pdf) Significance: Mar 25, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to address degraded control room air conditioning unit (Green) The inspectors identified a 10 CFR 50, Appendix B, Criterion XVI violation for the failure to take prompt corrective actions to address a degraded control room air conditioning unit (a condition adverse to quality). Since 1999, Grand Gulf engineers were aware that the Division I control room emergency air conditioning unit could not remove the required heat load under design basis conditions. The engineers failed to take prompt corrective measures to address the problem, because they did not have an accurate understanding of system requirements. The inspectors also identified that the licensee failed to properly address system operability on two occasions, as operability justifications were based on inaccurate or non-applicable information. This issue had cross-cutting aspects in the area of problem evaluation and prioritization. The failures to: 1) promptly correct a condition adverse to quality; and 2) properly evaluate equipment operability were performance deficiencies. The finding had more than minor significance because it affected the reactor safety mitigating systems objective to ensure the availability of systems that respond to initiating events. The finding was of very low risk significance because it was a design/qualification deficiency that did not result in a loss of function per Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual

1Q/2005 Inspection Findings - Grand Gulf 1 Page 2 of 4 Section on Resolution of Degraded and Nonconforming Conditions," Revision 1. Inspection Report# : 2005009(pdf) Significance: Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to retain safety related records for installation and testing of diesel generator starting air storage tank relief valves The inspectors identified a noncited violation of 10 CFR Part 50.71, "Maintenance of Records, Making of Reports," for failure of the licensee to retain safety related records relating to the periodic testing of the high pressure core spray emergency diesel generator starting air storage tank relief valves. This finding is more than minor because it is analogous to example 1.b of Appendix E of IMC 0612, "Power Reactor Inspection Reports," in that the safety related records were irretrievably lost. Using the Significance Determination Process Phase 1 worksheet, the inspectors determined the finding affected the mitigating systems cornerstone and was of very low safety significance because it did not represent an actual loss of system function. Inspection Report# : 2004005(pdf) Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Improper Valve Lineup Results in Isolation of RHR Pump Minimum Flow Line A self-revealing Green noncited violation of Technical Specification 5.4.1.a involved the failure of operators to comply with a valve lineup procedure prior to restoring the residual heat removal system to operation. This failure resulted in the isolation of the minimum flow line for the Train B residual heat removal pump, rendering one low pressure emergency core cooling system inoperable for 14 days, which violated the requirements of Technical Specification 3.5.1 prohibiting power operation with one low pressure emergency core cooling system out of service for greater than 7 days. This finding is greater than minor because it affected the configuration control and human performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. Using the Inspection Manual Chapter 0609 Significance Determination Process Phase 1 screening worksheet, this performance deficiency required a Phase 2 evaluation since it resulted in the actual loss of a single train for longer than its Technical Specification Allowed Outage Time. The Phase 2 and Phase 3 evaluations determined this finding to result in a core damage frequency change of less than 1.0E-6 and a change in Large Early Release Fraction of less than 1.0E-7. Therefore, the finding was considered to be of very low safety significance. Inspection Report# : 2004003(pdf) Barrier Integrity Significance: Jun 30, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Implement Reactor Startup Procedure Resulting in the Inadvertent Misalignment of the Control Rod Pattern The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a as a result of reactor operators failing to comply with an operating procedure used to establish a required rod pattern configuration during a reactor startup. This failure resulted in the reactor operators inadvertently withdrawing a control rod out of sequence. This finding is greater than minor because it involved the configuration control attribute (reactivity control) of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was of very low safety significance because it only affected the fuel barrier and not the reactor coolant system barrier. Inspection Report# : 2004003(pdf) Emergency Preparedness

1Q/2005 Inspection Findings - Grand Gulf 1 Page 3 of 4 Occupational Radiation Safety Significance: Mar 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to barricade and conspicuously post a high radiation area. The inspector identified a noncited violation of Technical Specification 5.7.1 because the licensee failed to barricade and conspicuously post a high radiation area. On March 16, 2005, during walkdowns of the reactor containment building 185-foot elevation, the inspector noted that a high radiation area posting in the reactor water clean-up sample sink area was not properly positioned across the access to the high radiation area. Radiation surveys taken in the area documented general area dose rates as high as 150 millirem per hour. This finding is greater than minor because it was associated with the cornerstone attribute (human performance) and affected the cornerstone objective because not posting a high radiation area with dose rates greater than 100 millirem per hour could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance. When licensee personnel exited the high radiation area and failed to ensure that the entrance was properly barricaded and conspicuously posted, their actions directly contributed to the finding. Inspection Report# : 2005002(pdf) Significance: Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to follow a source calibration procedure resulting in a worker receiving an unplanned, unintended dose. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1(a) for a worker who failed to follow a source calibration procedure and removed a lead attenuator while the radioactive source was in the up (exposed) position. As a result, the worker unintentionally exposed himself to a dose rate of approximately 330 millirem/hour, received an unplanned dose of one millirem and had the potential to receive additional unnecessary dose. This finding is greater than minor since it involves a worker's unplanned, unintended dose resulting from actions contrary to licensee procedures, which is associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and directly affects the cornerstone objective to ensure adequate protection of the worker's health and safety from exposure to radiation. The inspectors evaluated the finding using the Occupational Radiation Safety Significance Determination Process and determined it was of very low safety significance because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. Inspection Report# : 2004005(pdf) Significance: Jun 27, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Follow a Radiation Work Permit Requirement A self-revealing noncited violation of Technical Specification 5.4.1.a was evaluated for a worker who failed to follow a radiation work permit requirement. On March 15, 2004, a worker alarmed the personnel contamination monitors upon exiting the Radiologically Controlled Area because the individual had become contaminated. A follow-up survey of the work area identified contamination levels of up to 180,000 disintegrations per minute per 100 cm2 inside a drain pipe and 500,000 disintegrations per minute per 100 cm2 inside the valve housing. The licensee determined that the worker did not follow the radiation work permit requirement to contact Radiation Protection for approval before commencing cutting activities. This finding is greater than minor because it is associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding was of very low safety significance because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. Inspection Report# : 2004003(pdf) Public Radiation Safety Physical Protection

1Q/2005 Inspection Findings - Grand Gulf 1 Page 4 of 4 Physical Protection information not publicly available. Miscellaneous Significance: N/A Mar 25, 2005 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The team reviewed approximately 150 condition reports, apparent and root cause analyses, as well as other documents, to assess problem identification and resolution activities. Over the past two years (the assessment period) the team noted a few instances where problems were not properly identified, evaluated, prioritized or corrected but, overall, the licensee's processes were effective. Based on the interviews conducted, the team concluded that a positive safety-conscience work environment existed at Grand Gulf. The team determined that employees generally felt free to raise safety concerns to their supervision, the employee concerns program and the NRC. The team received a few isolated comments regarding: 1) a reluctance to use the site employee concerns program; 2) production pressure; and 3) the impact of staff reductions on work load and the ability to identify safety issues. Nonetheless, the interviewees all believed that potential safety issues were being addressed. The team determined that licensee management was aware of the perceptions and was taking action to address them. Inspection Report# : 2005009(pdf) Last modified : June 17, 2005

2Q/2005 Inspection Findings - Grand Gulf 1 Page 1 of 4 Grand Gulf 1 2Q/2005 Plant Inspection Findings Initiating Events Mitigating Systems Significance: May 12, 2005 Identified By: NRC Item Type: FIN Finding Inadequate Fire Drill Critique A finding was identified for fire brigade performance deficiencies that were not identified by the licensee during the drill critique. The deficiencies identified by the inspection team but not noted by the licensee's critique included not using lense inserts, using a fire hose that did not reach the fire properly, not maintaining a two-person rescue team, and not considering requesting offsite assistance. The licensee identified a number of additional performance deficiencies, and determined that performance during the May 10, 2005, unannounced fire drill was unsatisfactory. In accordance with the licensee's program, the individuals involved required remediation and the drill must be re-performed within 30 days. The licensee's incomplete assessment of fire brigade during the unannounced May 10, 2005, fire drill was a performance deficiency because the corrective action process would not have addressed the missed performance problems. This finding was more than minor because the Mitigating Systems cornerstone objective attribute to provide protection against external factors (fires) was affected. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," states that it excludes findings associated with the onsite manual fire brigade. Therefore, in accordance with Manual Chapter 0609, the safety significance was determined by regional management review. This review concluded that this finding was of very low safety significance because it reflected a training drill, rather than fire brigade fire performance during an actual fire. The fire brigade performance aspect of this finding affects the cross-cutting area of human performance. The incomplete drill evaluation aspect of this finding affects the crosscutting area of problem identification and resolution. This issue was entered into the licensee's corrective action program under Condition Report 2005-01872. Inspection Report# : 2005008(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: FIN Finding No Procedures for Implementing Two Repairs Needed to Achieve Cold Shutdown Following A CR Fire A finding was identified for not properly identifying repairs needed to achieve and maintain cold shutdown following a control room fire and documenting them in analyses and procedures. The team identified two repairs which were necessary in order to be able to achieve cold shutdown according to the licensee's alternate shutdown methodology. An alternate air supply was needed to maintain safety relief valves open during prolonged implementation of alternate shutdown cooling, and temporary instrumentation was needed to monitor reactor temperature and cooldown rate in the same mode. This issue was entered into the licensee's corrective action program under Condition Report 2005-02369. Failure to properly identify repairs needed to achieve and maintain cold shutdown following a control room fire and document them in analyses and procedures was a performance deficiency. This issue was more than minor because it affected the Mitigating Systems cornerstone attributes of protection from external factors (fire) and procedure quality. This finding was determined to have very low safety significance using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it involved an issue that only affected cold shutdown. Inspection Report# : 2005008(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: NCV NonCited Violation Alternative Shutdown Procedure Did Not Implement Safe Shutdown Analysis Assumption to Isolate Containment in a Timely Manner A noncited violation of 10 CFR 50, Appendix R, Section III.L.3 was identified for an inadequate alternative shutdown procedure. The team identified that Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 30, was not consistent with the safe shutdown analysis with respect to main steam isolation. The procedure did not require shutting the main steam isolation valves in a timely manner to prevent an excessive loss of reactor coolant in the event of a control room evacuation due to fire. Operators might not recognize the loss of coolant due to the limited indications available on the remote shutdown panel. This could result in loss of the reactor coolant makeup and decay

2Q/2005 Inspection Findings - Grand Gulf 1 Page 2 of 4 heat removal functions. The licensee promptly corrected the procedure and entered this issue in their corrective action program under Condition Report 2005-01865. Failure to assure that an important safe shutdown analysis assumption was translated into the alternative shutdown procedure was a performance deficiency. This issue was more than minor because it affected the Mitigating Systems cornerstone attributes of protection from external factors (fire) and procedure quality. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," Appendix F states that it excludes findings associated with control room evacuation. Therefore, in accordance with Manual Chapter 0609, the safety significance of this issue was determined by regional management review. This review concluded this finding was of very low safety significance because a licensee evaluation determined that the inventory lost from the reactor and containment through open main steam isolation valves following a control room fire would not affect low pressure injection for more than 24 hours. Also, during the initial stages, the inventory loss would be beneficial compared to promptly shutting the main steam isolation valves, since the steam would be removing significant heat that would otherwise have been retained in containment and would have to be removed through the remaining engineered safety features train. Therefore, additional assistance would be available from the technical support center and repair teams to help identify the problem and direct closure of the main steam isolation valves. Inspection Report# : 2005008(pdf) Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Operability Dertmination Procedure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for a failure to follow procedures that resulted in an inadequate operability determination. Specifically, operators failed to adequately implement the provisions of their operability determination to evaluate a degraded condition in the control room air conditioning system. This finding was greater than minor since it is associated with the equipment performance attribute of the mitigating systems cornerstone and directly affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process," the finding was of very low safety significance since: (1) it did not represent an actual loss of system safety function for the control room air conditioning system, (2) it did result in a loss of function for a single train of Technical Specification equipment, but for less than the Technical Specification allowed outage time, and (3) it did not represent a loss of function of non-technical specification risk significant equipment or screen as potentially risk significant due to a seismic, flooding or severe weather event. This finding has crosscutting aspects associated with human performance in that the control room operators failed to implement the operability determination procedure. Inspection Report# : 2005002(pdf) Significance: Mar 25, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to address degraded control room air conditioning unit (Green) The inspectors identified a 10 CFR 50, Appendix B, Criterion XVI violation for the failure to take prompt corrective actions to address a degraded control room air conditioning unit (a condition adverse to quality). Since 1999, Grand Gulf engineers were aware that the Division I control room emergency air conditioning unit could not remove the required heat load under design basis conditions. The engineers failed to take prompt corrective measures to address the problem, because they did not have an accurate understanding of system requirements. The inspectors also identified that the licensee failed to properly address system operability on two occasions, as operability justifications were based on inaccurate or non-applicable information. This issue had cross-cutting aspects in the area of problem evaluation and prioritization. The failures to: 1) promptly correct a condition adverse to quality; and 2) properly evaluate equipment operability were performance deficiencies. The finding had more than minor significance because it affected the reactor safety mitigating systems objective to ensure the availability of systems that respond to initiating events. The finding was of very low risk significance because it was a design/qualification deficiency that did not result in a loss of function per Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions," Revision 1. Inspection Report# : 2005009(pdf) Significance: Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to retain safety related records for installation and testing of diesel generator starting air storage tank relief valves The inspectors identified a noncited violation of 10 CFR Part 50.71, "Maintenance of Records, Making of Reports," for failure of the licensee to retain safety related records relating to the periodic testing of the high pressure core spray emergency diesel generator starting air storage tank relief valves. This finding is more than minor because it is analogous to example 1.b of Appendix E of IMC 0612, "Power Reactor Inspection Reports," in that the safety related records were irretrievably lost. Using the Significance Determination Process Phase 1 worksheet, the inspectors determined the finding affected the mitigating systems cornerstone and was of very low safety significance because it did not represent an actual loss of system function.

2Q/2005 Inspection Findings - Grand Gulf 1 Page 3 of 4 Inspection Report# : 2004005(pdf) Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Mar 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to barricade and conspicuously post a high radiation area. The inspector identified a noncited violation of Technical Specification 5.7.1 because the licensee failed to barricade and conspicuously post a high radiation area. On March 16, 2005, during walkdowns of the reactor containment building 185-foot elevation, the inspector noted that a high radiation area posting in the reactor water clean-up sample sink area was not properly positioned across the access to the high radiation area. Radiation surveys taken in the area documented general area dose rates as high as 150 millirem per hour. This finding is greater than minor because it was associated with the cornerstone attribute (human performance) and affected the cornerstone objective because not posting a high radiation area with dose rates greater than 100 millirem per hour could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance. When licensee personnel exited the high radiation area and failed to ensure that the entrance was properly barricaded and conspicuously posted, their actions directly contributed to the finding. Inspection Report# : 2005002(pdf) Significance: Dec 31, 2004 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to follow a source calibration procedure resulting in a worker receiving an unplanned, unintended dose. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1(a) for a worker who failed to follow a source calibration procedure and removed a lead attenuator while the radioactive source was in the up (exposed) position. As a result, the worker unintentionally exposed himself to a dose rate of approximately 330 millirem/hour, received an unplanned dose of one millirem and had the potential to receive additional unnecessary dose. This finding is greater than minor since it involves a worker's unplanned, unintended dose resulting from actions contrary to licensee procedures, which is associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and directly affects the cornerstone objective to ensure adequate protection of the worker's health and safety from exposure to radiation. The inspectors evaluated the finding using the Occupational Radiation Safety Significance Determination Process and determined it was of very low safety significance because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. Inspection Report# : 2004005(pdf) Public Radiation Safety Physical Protection Physical Protection information not publicly available.

2Q/2005 Inspection Findings - Grand Gulf 1 Page 4 of 4 Miscellaneous Significance: N/A Mar 25, 2005 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The team reviewed approximately 150 condition reports, apparent and root cause analyses, as well as other documents, to assess problem identification and resolution activities. Over the past two years (the assessment period) the team noted a few instances where problems were not properly identified, evaluated, prioritized or corrected but, overall, the licensee's processes were effective. Based on the interviews conducted, the team concluded that a positive safety-conscience work environment existed at Grand Gulf. The team determined that employees generally felt free to raise safety concerns to their supervision, the employee concerns program and the NRC. The team received a few isolated comments regarding: 1) a reluctance to use the site employee concerns program; 2) production pressure; and 3) the impact of staff reductions on work load and the ability to identify safety issues. Nonetheless, the interviewees all believed that potential safety issues were being addressed. The team determined that licensee management was aware of the perceptions and was taking action to address them. Inspection Report# : 2005009(pdf) Last modified : August 24, 2005

3Q/2005 Inspection Findings - Grand Gulf 1 Page 1 of 5 Grand Gulf 1 3Q/2005 Plant Inspection Findings Initiating Events Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: FIN Finding Inadvertent Plant Service Water Pump Trip A Green self-revealing finding was identified for the inadvertent trip of a plant service water pump due to an inadequate procedure. The procedure failed to perform its stated purpose to verify the operation of a service water pump support system. The licensee entered this performance deficiency in their corrective action program for resolution. This finding is more than minor since it affected the configuration control and human performance attributes of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of the finding is related to the resources aspect of the cross-cutting area of human performance. Inspection Report# : 2005004(pdf) Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: FIN Finding Improper Maintenance Results in Partial Loss of Component Cooling Water A self-revealing Green finding was reviewed for the failure of a newly installed corrosion monitor probe that resulted in a leak in the component cooling water system. Licensee personnel used an inadequate procedure to install the probe and therefore failed to verify the pressure retaining capability of the probe prior to installation. The licensee entered this performance deficiency in their corrective action program. This finding is more than minor since it affected the design control attribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of the finding is related to the resources aspect of the cross-cutting area of human performance. Inspection Report# : 2005004(pdf) Mitigating Systems Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Disabling of Diesel Generator Alarms due to Failure to Follow Procedure The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1 for a failure to follow procedure that resulted in the disabling of required supervisory alarms on the Division II emergency diesel generator. Specifically, operators failed to reset the alarm panel following routine testing. The licensee entered this performance deficiency into their corrective action program. This finding is more than minor since the disabling of required alarm functions for the emergency diesel generators could become a more significant safety concern if left uncorrected. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not result in an actual loss of the safety function. The cause of the finding is related to the personnel aspect of the cross-cutting area of human performance. Inspection Report# : 2005004(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: NCV NonCited Violation Alternative Shutdown Procedure Did Not Implement Safe Shutdown Analysis Assumption to Isolate Containment in a Timely Manner A noncited violation of 10 CFR 50, Appendix R, Section III.L.3 was identified for an inadequate alternative shutdown procedure. The team

3Q/2005 Inspection Findings - Grand Gulf 1 Page 2 of 5 identified that Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 30, was not consistent with the safe shutdown analysis with respect to main steam isolation. The procedure did not require shutting the main steam isolation valves in a timely manner to prevent an excessive loss of reactor coolant in the event of a control room evacuation due to fire. Operators might not recognize the loss of coolant due to the limited indications available on the remote shutdown panel. This could result in loss of the reactor coolant makeup and decay heat removal functions. The licensee promptly corrected the procedure and entered this issue in their corrective action program under Condition Report 2005-01865. Failure to assure that an important safe shutdown analysis assumption was translated into the alternative shutdown procedure was a performance deficiency. This issue was more than minor because it affected the Mitigating Systems cornerstone attributes of protection from external factors (fire) and procedure quality. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," Appendix F states that it excludes findings associated with control room evacuation. Therefore, in accordance with Manual Chapter 0609, the safety significance of this issue was determined by regional management review. This review concluded this finding was of very low safety significance because a licensee evaluation determined that the inventory lost from the reactor and containment through open main steam isolation valves following a control room fire would not affect low pressure injection for more than 24 hours. Also, during the initial stages, the inventory loss would be beneficial compared to promptly shutting the main steam isolation valves, since the steam would be removing significant heat that would otherwise have been retained in containment and would have to be removed through the remaining engineered safety features train. Therefore, additional assistance would be available from the technical support center and repair teams to help identify the problem and direct closure of the main steam isolation valves. Inspection Report# : 2005008(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: FIN Finding Inadequate Fire Drill Critique A finding was identified for fire brigade performance deficiencies that were not identified by the licensee during the drill critique. The deficiencies identified by the inspection team but not noted by the licensee's critique included not using lense inserts, using a fire hose that did not reach the fire properly, not maintaining a two-person rescue team, and not considering requesting offsite assistance. The licensee identified a number of additional performance deficiencies, and determined that performance during the May 10, 2005, unannounced fire drill was unsatisfactory. In accordance with the licensee's program, the individuals involved required remediation and the drill must be re-performed within 30 days. The licensee's incomplete assessment of fire brigade during the unannounced May 10, 2005, fire drill was a performance deficiency because the corrective action process would not have addressed the missed performance problems. This finding was more than minor because the Mitigating Systems cornerstone objective attribute to provide protection against external factors (fires) was affected. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," states that it excludes findings associated with the onsite manual fire brigade. Therefore, in accordance with Manual Chapter 0609, the safety significance was determined by regional management review. This review concluded that this finding was of very low safety significance because it reflected a training drill, rather than fire brigade fire performance during an actual fire. The fire brigade performance aspect of this finding affects the cross-cutting area of human performance. The incomplete drill evaluation aspect of this finding affects the crosscutting area of problem identification and resolution. This issue was entered into the licensee's corrective action program under Condition Report 2005-01872. Inspection Report# : 2005008(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: FIN Finding No Procedures for Implementing Two Repairs Needed to Achieve Cold Shutdown Following A CR Fire A finding was identified for not properly identifying repairs needed to achieve and maintain cold shutdown following a control room fire and documenting them in analyses and procedures. The team identified two repairs which were necessary in order to be able to achieve cold shutdown according to the licensee's alternate shutdown methodology. An alternate air supply was needed to maintain safety relief valves open during prolonged implementation of alternate shutdown cooling, and temporary instrumentation was needed to monitor reactor temperature and cooldown rate in the same mode. This issue was entered into the licensee's corrective action program under Condition Report 2005-02369. Failure to properly identify repairs needed to achieve and maintain cold shutdown following a control room fire and document them in analyses and procedures was a performance deficiency. This issue was more than minor because it affected the Mitigating Systems cornerstone attributes of protection from external factors (fire) and procedure quality. This finding was determined to have very low safety significance using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it involved an issue that only affected cold shutdown. Inspection Report# : 2005008(pdf) Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Operability Dertmination Procedure

3Q/2005 Inspection Findings - Grand Gulf 1 Page 3 of 5 The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for a failure to follow procedures that resulted in an inadequate operability determination. Specifically, operators failed to adequately implement the provisions of their operability determination to evaluate a degraded condition in the control room air conditioning system. This finding was greater than minor since it is associated with the equipment performance attribute of the mitigating systems cornerstone and directly affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process," the finding was of very low safety significance since: (1) it did not represent an actual loss of system safety function for the control room air conditioning system, (2) it did result in a loss of function for a single train of Technical Specification equipment, but for less than the Technical Specification allowed outage time, and (3) it did not represent a loss of function of non-technical specification risk significant equipment or screen as potentially risk significant due to a seismic, flooding or severe weather event. This finding has crosscutting aspects associated with human performance in that the control room operators failed to implement the operability determination procedure. Inspection Report# : 2005002(pdf) Significance: Mar 25, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to address degraded control room air conditioning unit (Green) The inspectors identified a 10 CFR 50, Appendix B, Criterion XVI violation for the failure to take prompt corrective actions to address a degraded control room air conditioning unit (a condition adverse to quality). Since 1999, Grand Gulf engineers were aware that the Division I control room emergency air conditioning unit could not remove the required heat load under design basis conditions. The engineers failed to take prompt corrective measures to address the problem, because they did not have an accurate understanding of system requirements. The inspectors also identified that the licensee failed to properly address system operability on two occasions, as operability justifications were based on inaccurate or non-applicable information. This issue had cross-cutting aspects in the area of problem evaluation and prioritization. The failures to: 1) promptly correct a condition adverse to quality; and 2) properly evaluate equipment operability were performance deficiencies. The finding had more than minor significance because it affected the reactor safety mitigating systems objective to ensure the availability of systems that respond to initiating events. The finding was of very low risk significance because it was a design/qualification deficiency that did not result in a loss of function per Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions," Revision 1. Inspection Report# : 2005009(pdf) Significance: Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to retain safety related records for installation and testing of diesel generator starting air storage tank relief valves The inspectors identified a noncited violation of 10 CFR Part 50.71, "Maintenance of Records, Making of Reports," for failure of the licensee to retain safety related records relating to the periodic testing of the high pressure core spray emergency diesel generator starting air storage tank relief valves. This finding is more than minor because it is analogous to example 1.b of Appendix E of IMC 0612, "Power Reactor Inspection Reports," in that the safety related records were irretrievably lost. Using the Significance Determination Process Phase 1 worksheet, the inspectors determined the finding affected the mitigating systems cornerstone and was of very low safety significance because it did not represent an actual loss of system function. Inspection Report# : 2004005(pdf) Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Sep 30, 2005 Identified By: Self-Revealing

3Q/2005 Inspection Findings - Grand Gulf 1 Page 4 of 5 Item Type: NCV NonCited Violation Failure to Control a High Radiation Area with Dose Rates Greater than One Rem per Hour The inspector reviewed a self-revealing non-cited violation of Technical Specification 5.7.3 because the licensee failed to control a high radiation area with dose rates greater than 1,000 millirem per hour. Specifically, on September 22, 2005, a radiation worker was performing a visual inspection of a low pressure coolant injection pipe penetration in the drywell. The worker climbed three feet above the floor elevation, at which time the worker's electronic dosimeter alarmed with peak dose rate of 582 millirem per hour. Radiation protection personnel performed a survey of the area and determined that dose rates were as high as 1,200 millirem per hour at one foot from the low pressure coolant injection pipe. This finding was entered into the licensee's corrective action program. This finding is greater than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involves the potential for a worker's unplanned or unintended dose resulting from actions contrary to technical specifications. When processed through the Occupational Radiation Safety Significance Determination Process, the finding is of very low safety significance because it did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2005004(pdf) Significance: Mar 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to barricade and conspicuously post a high radiation area. The inspector identified a noncited violation of Technical Specification 5.7.1 because the licensee failed to barricade and conspicuously post a high radiation area. On March 16, 2005, during walkdowns of the reactor containment building 185-foot elevation, the inspector noted that a high radiation area posting in the reactor water clean-up sample sink area was not properly positioned across the access to the high radiation area. Radiation surveys taken in the area documented general area dose rates as high as 150 millirem per hour. This finding is greater than minor because it was associated with the cornerstone attribute (human performance) and affected the cornerstone objective because not posting a high radiation area with dose rates greater than 100 millirem per hour could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance. When licensee personnel exited the high radiation area and failed to ensure that the entrance was properly barricaded and conspicuously posted, their actions directly contributed to the finding. Inspection Report# : 2005002(pdf) Significance: Dec 31, 2004 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to follow a source calibration procedure resulting in a worker receiving an unplanned, unintended dose. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1(a) for a worker who failed to follow a source calibration procedure and removed a lead attenuator while the radioactive source was in the up (exposed) position. As a result, the worker unintentionally exposed himself to a dose rate of approximately 330 millirem/hour, received an unplanned dose of one millirem and had the potential to receive additional unnecessary dose. This finding is greater than minor since it involves a worker's unplanned, unintended dose resulting from actions contrary to licensee procedures, which is associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and directly affects the cornerstone objective to ensure adequate protection of the worker's health and safety from exposure to radiation. The inspectors evaluated the finding using the Occupational Radiation Safety Significance Determination Process and determined it was of very low safety significance because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. Inspection Report# : 2004005(pdf) Public Radiation Safety Physical Protection Physical Protection information not publicly available.

3Q/2005 Inspection Findings - Grand Gulf 1 Page 5 of 5 Miscellaneous Significance: N/A Mar 25, 2005 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The team reviewed approximately 150 condition reports, apparent and root cause analyses, as well as other documents, to assess problem identification and resolution activities. Over the past two years (the assessment period) the team noted a few instances where problems were not properly identified, evaluated, prioritized or corrected but, overall, the licensee's processes were effective. Based on the interviews conducted, the team concluded that a positive safety-conscience work environment existed at Grand Gulf. The team determined that employees generally felt free to raise safety concerns to their supervision, the employee concerns program and the NRC. The team received a few isolated comments regarding: 1) a reluctance to use the site employee concerns program; 2) production pressure; and 3) the impact of staff reductions on work load and the ability to identify safety issues. Nonetheless, the interviewees all believed that potential safety issues were being addressed. The team determined that licensee management was aware of the perceptions and was taking action to address them. Inspection Report# : 2005009(pdf) Last modified : November 30, 2005

4Q/2005 Inspection Findings - Grand Gulf 1 Page 1 of 5 Grand Gulf 1 4Q/2005 Plant Inspection Findings Initiating Events Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: FIN Finding Inadvertent Plant Service Water Pump Trip A Green self-revealing finding was identified for the inadvertent trip of a plant service water pump due to a failure to follow procedure. In addition, the procedure did not meet its stated purpose to verify the operation of a service water pump support system, specifically the well level indication system. The licensee entered this performance deficiency in their corrective action program for resolution. This finding is more than minor since it affected the configuration control and human performance attributes of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. This finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: FIN Finding Improper Maintenance Results in Partial Loss of Component Cooling Water A Green self-revealing finding was reviewed involving the failure of a newly installed corrosion monitor probe that resulted in a leak in the component cooling water system. Licensee personnel used an inadequate procedure to install the probe and therefore failed to verify the pressure retaining capability of the probe prior to installation. The licensee entered this performance deficiency in their corrective action program for resolution. This finding is more than minor since it affected the design control attribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Mitigating Systems Significance: Oct 13, 2005 Identified By: NRC Item Type: NCV NonCited Violation Foreign Material in the Drywell The inspectors identified a Green noncited violation (NCV) of TS 5.4.1(a) for the failure of licensee personnel to perform an adequate drywell closeout inspection for foreign material. On October 13, 2005, licensee personnel failed to follow Integrated Operating Instruction 3-1-01-1, "Cold Shutdown to Minimum Generator Load," Attachment II, steps 2, 16, 21, and 34 of the drywell closeout sheet. The inspectors conducted a general inspection of the drywell and discovered approximately 50 foreign material items totaling a volume of approximately one and a half cubic feet in the drywell floor area. This foreign material included plastic wrappings and tie-wraps, articles of protective clothing, loose paper, metal objects and other miscellaneous material. This issue was entered into the licensee's corrective action program as CR-GGN-2006-00236. The finding is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding is determined to have very low safety significance because there was no loss of safety function of the emergency core cooling system suction strainers. The cause of the finding is related to the crosscutting element of human performance in that licensee personnel did not follow the drywell closeout procedure. Inspection Report# : 2005005(pdf)

4Q/2005 Inspection Findings - Grand Gulf 1 Page 2 of 5 Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Disabling of Diesel Generator Alarms due to Failure to Follow Procedure The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1 involving a failure to follow procedure that resulted in the disabling of required supervisory alarms on the Division II emergency diesel generator. Specifically, operators failed to reset the alarm panel following routine testing. The licensee entered this performance deficiency into their corrective action program. This finding is more than minor since the disabling of required alarm functions for the emergency diesel generators could become a more significant safety concern if left uncorrected. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not result in an actual loss of the safety function. This finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: NCV NonCited Violation Alternative Shutdown Procedure Did Not Implement Safe Shutdown Analysis Assumption to Isolate Containment in a Timely Manner A noncited violation of 10 CFR 50, Appendix R, Section III.L.3 was identified for an inadequate alternative shutdown procedure. The team identified that Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 30, was not consistent with the safe shutdown analysis with respect to main steam isolation. The procedure did not require shutting the main steam isolation valves in a timely manner to prevent an excessive loss of reactor coolant in the event of a control room evacuation due to fire. Operators might not recognize the loss of coolant due to the limited indications available on the remote shutdown panel. This could result in loss of the reactor coolant makeup and decay heat removal functions. The licensee promptly corrected the procedure and entered this issue in their corrective action program under Condition Report 2005-01865. Failure to assure that an important safe shutdown analysis assumption was translated into the alternative shutdown procedure was a performance deficiency. This issue was more than minor because it affected the Mitigating Systems cornerstone attributes of protection from external factors (fire) and procedure quality. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," Appendix F states that it excludes findings associated with control room evacuation. Therefore, in accordance with Manual Chapter 0609, the safety significance of this issue was determined by regional management review. This review concluded this finding was of very low safety significance because a licensee evaluation determined that the inventory lost from the reactor and containment through open main steam isolation valves following a control room fire would not affect low pressure injection for more than 24 hours. Also, during the initial stages, the inventory loss would be beneficial compared to promptly shutting the main steam isolation valves, since the steam would be removing significant heat that would otherwise have been retained in containment and would have to be removed through the remaining engineered safety features train. Therefore, additional assistance would be available from the technical support center and repair teams to help identify the problem and direct closure of the main steam isolation valves. Inspection Report# : 2005008(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: FIN Finding Inadequate Fire Drill Critique A finding was identified for fire brigade performance deficiencies that were not identified by the licensee during the drill critique. The deficiencies identified by the inspection team but not noted by the licensee's critique included not using lense inserts, using a fire hose that did not reach the fire properly, not maintaining a two-person rescue team, and not considering requesting offsite assistance. The licensee identified a number of additional performance deficiencies, and determined that performance during the May 10, 2005, unannounced fire drill was unsatisfactory. In accordance with the licensee's program, the individuals involved required remediation and the drill must be re-performed within 30 days. The licensee's incomplete assessment of fire brigade during the unannounced May 10, 2005, fire drill was a performance deficiency because the corrective action process would not have addressed the missed performance problems. This finding was more than minor because the Mitigating Systems cornerstone objective attribute to provide protection against external factors (fires) was affected. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," states that it excludes findings associated with the onsite manual fire brigade. Therefore, in accordance with Manual Chapter 0609, the safety significance was determined by regional management review. This review concluded that this finding was of very low safety significance because it reflected a training drill, rather than fire brigade fire performance during an actual fire. The fire brigade performance aspect of this finding affects the cross-cutting area of human performance. The incomplete drill evaluation aspect of this finding affects the crosscutting area of problem identification and resolution. This issue was entered into the licensee's corrective action program under Condition Report 2005-01872. Inspection Report# : 2005008(pdf) Significance: May 12, 2005 Identified By: NRC

4Q/2005 Inspection Findings - Grand Gulf 1 Page 3 of 5 Item Type: FIN Finding No Procedures for Implementing Two Repairs Needed to Achieve Cold Shutdown Following A CR Fire A finding was identified for not properly identifying repairs needed to achieve and maintain cold shutdown following a control room fire and documenting them in analyses and procedures. The team identified two repairs which were necessary in order to be able to achieve cold shutdown according to the licensee's alternate shutdown methodology. An alternate air supply was needed to maintain safety relief valves open during prolonged implementation of alternate shutdown cooling, and temporary instrumentation was needed to monitor reactor temperature and cooldown rate in the same mode. This issue was entered into the licensee's corrective action program under Condition Report 2005-02369. Failure to properly identify repairs needed to achieve and maintain cold shutdown following a control room fire and document them in analyses and procedures was a performance deficiency. This issue was more than minor because it affected the Mitigating Systems cornerstone attributes of protection from external factors (fire) and procedure quality. This finding was determined to have very low safety significance using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it involved an issue that only affected cold shutdown. Inspection Report# : 2005008(pdf) Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Operability Dertmination Procedure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for a failure to follow procedures that resulted in an inadequate operability determination. Specifically, operators failed to adequately implement the provisions of their operability determination to evaluate a degraded condition in the control room air conditioning system. This finding was greater than minor since it is associated with the equipment performance attribute of the mitigating systems cornerstone and directly affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process," the finding was of very low safety significance since: (1) it did not represent an actual loss of system safety function for the control room air conditioning system, (2) it did result in a loss of function for a single train of Technical Specification equipment, but for less than the Technical Specification allowed outage time, and (3) it did not represent a loss of function of non-technical specification risk significant equipment or screen as potentially risk significant due to a seismic, flooding or severe weather event. This finding has crosscutting aspects associated with human performance in that the control room operators failed to implement the operability determination procedure. Inspection Report# : 2005002(pdf) Significance: Mar 25, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate corrective actions to address degraded control room air conditioning unit (Green) The inspectors identified a 10 CFR 50, Appendix B, Criterion XVI violation for the failure to take prompt corrective actions to address a degraded control room air conditioning unit (a condition adverse to quality). Since 1999, Grand Gulf engineers were aware that the Division I control room emergency air conditioning unit could not remove the required heat load under design basis conditions. The engineers failed to take prompt corrective measures to address the problem, because they did not have an accurate understanding of system requirements. The inspectors also identified that the licensee failed to properly address system operability on two occasions, as operability justifications were based on inaccurate or non-applicable information. This issue had cross-cutting aspects in the area of problem evaluation and prioritization. The failures to: 1) promptly correct a condition adverse to quality; and 2) properly evaluate equipment operability were performance deficiencies. The finding had more than minor significance because it affected the reactor safety mitigating systems objective to ensure the availability of systems that respond to initiating events. The finding was of very low risk significance because it was a design/qualification deficiency that did not result in a loss of function per Generic Letter 91-18, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions," Revision 1. Inspection Report# : 2005009(pdf) Barrier Integrity Emergency Preparedness

4Q/2005 Inspection Findings - Grand Gulf 1 Page 4 of 5 Occupational Radiation Safety Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Control a High Radiation Area with Dose Rates Greater than One Rem per Hour The inspector reviewed a self-revealing non-cited violation of Technical Specification 5.7.3 involving the licensee's failure to control a high radiation area with dose rates greater than 1,000 millirem per hour. Specifically, on September 22, 2005, a radiation worker was performing a visual inspection of a low pressure coolant injection pipe penetration in the drywell. The worker climbed three feet above the floor elevation, at which time the worker's electronic dosimeter alarmed with peak dose rate of 582 millirem per hour. Radiation protection personnel performed a survey of the area and determined that dose rates were as high as 1,200 millirem per hour at one foot from the low pressure coolant injection pipe. This finding was entered into the licensee's corrective action program. This finding is greater than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involves the potential for a worker's unplanned or unintended dose resulting from actions contrary to technical specifications. When processed through the Occupational Radiation Safety Significance Determination Process, the finding is of very low safety significance because it did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2005004(pdf) Significance: Mar 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to barricade and conspicuously post a high radiation area. The inspector identified a noncited violation of Technical Specification 5.7.1 because the licensee failed to barricade and conspicuously post a high radiation area. On March 16, 2005, during walkdowns of the reactor containment building 185-foot elevation, the inspector noted that a high radiation area posting in the reactor water clean-up sample sink area was not properly positioned across the access to the high radiation area. Radiation surveys taken in the area documented general area dose rates as high as 150 millirem per hour. This finding is greater than minor because it was associated with the cornerstone attribute (human performance) and affected the cornerstone objective because not posting a high radiation area with dose rates greater than 100 millirem per hour could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined that the finding was of very low safety significance because it did not involve (1) ALARA planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. Additionally, this finding had crosscutting aspects associated with human performance. When licensee personnel exited the high radiation area and failed to ensure that the entrance was properly barricaded and conspicuously posted, their actions directly contributed to the finding. Inspection Report# : 2005002(pdf) Public Radiation Safety Physical Protection Physical Protection information not publicly available. Miscellaneous Significance: N/A Mar 25, 2005 Identified By: NRC Item Type: FIN Finding Problem Identification and Resolution The team reviewed approximately 150 condition reports, apparent and root cause analyses, as well as other documents, to assess problem identification and resolution activities. Over the past two years (the assessment period) the team noted a few instances where problems were not properly identified, evaluated, prioritized or corrected but, overall, the licensee's processes were effective. Based on the interviews conducted, the team concluded that a positive safety-conscience work environment existed at Grand Gulf. The team

4Q/2005 Inspection Findings - Grand Gulf 1 Page 5 of 5 determined that employees generally felt free to raise safety concerns to their supervision, the employee concerns program and the NRC. The team received a few isolated comments regarding: 1) a reluctance to use the site employee concerns program; 2) production pressure; and 3) the impact of staff reductions on work load and the ability to identify safety issues. Nonetheless, the interviewees all believed that potential safety issues were being addressed. The team determined that licensee management was aware of the perceptions and was taking action to address them. Inspection Report# : 2005009(pdf) Last modified : March 03, 2006

1Q/2006 Inspection Findings - Grand Gulf 1 Page 1 of 5 Grand Gulf 1 1Q/2006 Plant Inspection Findings Initiating Events Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment The inspectors identified two examples of a noncited violation of 10 CFR 50.65, Maintenance Rule, for failing to include maintenance that could increase the likelihood of an initiating event in the plant risk assessment. On February 2, 2006 and again on March 28, 2006, the licensees risk assessment did not include maintenance activities that increased the likelihood of a reactor scram. The licensee entered this into their corrective action program as Condition Reports CR-GGN-2006-1041 and CR-GGN-2006-1277. This finding is more than minor since the maintenance that was performed increased the likelihood of an initiating event. Using Inspection Manual Chapter 0609 Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is of very low safety significance since in both cases the change in incremental core damage probability and incremental large early release probability were less than 1E-6 and 1E-7, respectively. This finding has human performance crosscutting aspects because the inadequate risk assessments were due to personnel error. Inspection Report# : 2006002(pdf) Significance: Mar 31, 2006 Identified By: Self-Revealing Item Type: FIN Finding Plant Service Water Leak During Excavation The inspectors reviewed a self-revealing finding for a failure to follow procedure that resulted in a significant plant service water header leak. The licensee failed to perform an adequate review of documents to identify potential hazards as required by Procedure EN-S-112, Trenching, Excavation and Ground Penetrating Activities, Revision 2. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-0219. This finding is more than minor since it was associated with the human performance attribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of this finding has human performance cross-cutting aspects associated with a failure to follow procedures. Inspection Report# : 2006002(pdf) Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: FIN Finding Inadvertent Plant Service Water Pump Trip A Green self-revealing finding was identified for the inadvertent trip of a plant service water pump due to a failure to follow procedure. In addition, the procedure did not meet its stated purpose to verify the operation of a service water pump support system, specifically the well level indication system. The licensee entered this performance deficiency in their corrective action program for resolution. This finding is more than minor since it affected the configuration control and human performance attributes of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. This finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: FIN Finding Improper Maintenance Results in Partial Loss of Component Cooling Water A Green self-revealing finding was reviewed involving the failure of a newly installed corrosion monitor probe that resulted in a leak in the component cooling water system. Licensee personnel used an inadequate procedure to install the probe and therefore failed to verify the pressure retaining capability of the probe prior to installation. The licensee entered this performance deficiency in their corrective action program for resolution.

1Q/2006 Inspection Findings - Grand Gulf 1 Page 2 of 5 This finding is more than minor since it affected the design control attribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Mitigating Systems Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Associated with Condensate Storage Tank Level Instrumentation The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to take prompt corrective actions to address a design deficiency involving condensate storage tank level instrumentation. The licensee identified the design deficiency on April 30, 1999, and issued compensatory actions for the operators to manually transfer high pressure core spray and reactor core isolation cooling from the condensate storage tank to the suppression pool in the event of failure of the tank. The licensee corrected the design deficiency on December 8, 2005. The licensee entered this issue in their corrective action program as CR-GGN-2006-1096. This finding is more than minor because it affected the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was of very low safety significance because it was a design deficiency that did not result in a loss of operability. This finding had cross-cutting aspects associated with problem identification and resolution in that station personnel did not implement corrective actions in a timely manner. Inspection Report# : 2006002(pdf) Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Alternate Shutdown Cooling Mode Not Properly Implemented in Alternate Shutdown Procedure A noncited violation of Grand Gulf License Condition 2.C(41), Fire Protection Program, was identified for failure to have an alternative shutdown procedure to restore power following a control room evacuation with loss of offsite power that was independent of the control room. Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 30, required operators to use System Operating Instruction 04-1-01-P75-1, "Standby Diesel Generator System," Revision 67, to locally start the Division 1 EDG in the event that offsite power was not available. However, this procedure included steps to shut the diesel generator output breaker from inside the control room, which would not be possible once the control room was evacuated. The licensee promptly corrected the procedure and entered this issue in their corrective action program under Condition Report 2005-01865. This issue had human performance cross-cutting aspects associated with an inadequate procedure. This issue was more than minor because it affected the Mitigating Systems Cornerstone objective under the procedure quality and protection from external factors attributes. This issue was categorized as a Post-fire Safe Shutdown issue associated with response procedure quality. The degradation rating was determined to be Low because operator experience and familiarity with performing the required response actions were adequate to overcome the procedure deficiency. Therefore, this issue screened as having very low safety significance (Green) in Phase 1 of the Fire Protection Significance Determination Process (Manual Chapter 0609, Appendix F). Inspection Report# : 2006002(pdf) Significance: Oct 13, 2005 Identified By: NRC Item Type: NCV NonCited Violation Foreign Material in the Drywell The inspectors identified a Green noncited violation (NCV) of TS 5.4.1(a) for the failure of licensee personnel to perform an adequate drywell closeout inspection for foreign material. On October 13, 2005, licensee personnel failed to follow Integrated Operating Instruction 3-1-01-1, "Cold Shutdown to Minimum Generator Load," Attachment II, steps 2, 16, 21, and 34 of the drywell closeout sheet. The inspectors conducted a general inspection of the drywell and discovered approximately 50 foreign material items totaling a volume of approximately one and a half cubic feet in the drywell floor area. This foreign material included plastic wrappings and tie-wraps, articles of protective clothing, loose paper, metal objects and other miscellaneous material. This issue was entered into the licensee's corrective action program as CR-GGN-2006-00236. The finding is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding is determined to have very low safety significance because there was no loss of safety function of the emergency core cooling system suction strainers. The cause of the finding is related to the crosscutting element of human performance in that licensee personnel did not follow the drywell closeout procedure.

1Q/2006 Inspection Findings - Grand Gulf 1 Page 3 of 5 Inspection Report# : 2005005(pdf) Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Disabling of Diesel Generator Alarms due to Failure to Follow Procedure The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1 involving a failure to follow procedure that resulted in the disabling of required supervisory alarms on the Division II emergency diesel generator. Specifically, operators failed to reset the alarm panel following routine testing. The licensee entered this performance deficiency into their corrective action program. This finding is more than minor since the disabling of required alarm functions for the emergency diesel generators could become a more significant safety concern if left uncorrected. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not result in an actual loss of the safety function. This finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: NCV NonCited Violation Alternative Shutdown Procedure Did Not Implement Safe Shutdown Analysis Assumption to Isolate Containment in a Timely Manner A noncited violation of 10 CFR 50, Appendix R, Section III.L.3 was identified for an inadequate alternative shutdown procedure. The team identified that Procedure 05-1-02-II-1, "Shutdown from the Remote Shutdown Panel," Revision 30, was not consistent with the safe shutdown analysis with respect to main steam isolation. The procedure did not require shutting the main steam isolation valves in a timely manner to prevent an excessive loss of reactor coolant in the event of a control room evacuation due to fire. Operators might not recognize the loss of coolant due to the limited indications available on the remote shutdown panel. This could result in loss of the reactor coolant makeup and decay heat removal functions. The licensee promptly corrected the procedure and entered this issue in their corrective action program under Condition Report 2005-01865. Failure to assure that an important safe shutdown analysis assumption was translated into the alternative shutdown procedure was a performance deficiency. This issue was more than minor because it affected the Mitigating Systems cornerstone attributes of protection from external factors (fire) and procedure quality. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," Appendix F states that it excludes findings associated with control room evacuation. Therefore, in accordance with Manual Chapter 0609, the safety significance of this issue was determined by regional management review. This review concluded this finding was of very low safety significance because a licensee evaluation determined that the inventory lost from the reactor and containment through open main steam isolation valves following a control room fire would not affect low pressure injection for more than 24 hours. Also, during the initial stages, the inventory loss would be beneficial compared to promptly shutting the main steam isolation valves, since the steam would be removing significant heat that would otherwise have been retained in containment and would have to be removed through the remaining engineered safety features train. Therefore, additional assistance would be available from the technical support center and repair teams to help identify the problem and direct closure of the main steam isolation valves. Inspection Report# : 2005008(pdf) Significance: May 12, 2005 Identified By: NRC Item Type: FIN Finding Inadequate Fire Drill Critique A finding was identified for fire brigade performance deficiencies that were not identified by the licensee during the drill critique. The deficiencies identified by the inspection team but not noted by the licensee's critique included not using lense inserts, using a fire hose that did not reach the fire properly, not maintaining a two-person rescue team, and not considering requesting offsite assistance. The licensee identified a number of additional performance deficiencies, and determined that performance during the May 10, 2005, unannounced fire drill was unsatisfactory. In accordance with the licensee's program, the individuals involved required remediation and the drill must be re-performed within 30 days. The licensee's incomplete assessment of fire brigade during the unannounced May 10, 2005, fire drill was a performance deficiency because the corrective action process would not have addressed the missed performance problems. This finding was more than minor because the Mitigating Systems cornerstone objective attribute to provide protection against external factors (fires) was affected. Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," states that it excludes findings associated with the onsite manual fire brigade. Therefore, in accordance with Manual Chapter 0609, the safety significance was determined by regional management review. This review concluded that this finding was of very low safety significance because it reflected a training drill, rather than fire brigade fire performance during an actual fire. The fire brigade performance aspect of this finding affects the cross-cutting area of human performance. The incomplete drill evaluation aspect of this finding affects the crosscutting area of problem identification and resolution. This issue was entered into the licensee's corrective action program under Condition Report 2005-01872. Inspection Report# : 2005008(pdf)

1Q/2006 Inspection Findings - Grand Gulf 1 Page 4 of 5 Significance: May 12, 2005 Identified By: NRC Item Type: FIN Finding No Procedures for Implementing Two Repairs Needed to Achieve Cold Shutdown Following A CR Fire A finding was identified for not properly identifying repairs needed to achieve and maintain cold shutdown following a control room fire and documenting them in analyses and procedures. The team identified two repairs which were necessary in order to be able to achieve cold shutdown according to the licensee's alternate shutdown methodology. An alternate air supply was needed to maintain safety relief valves open during prolonged implementation of alternate shutdown cooling, and temporary instrumentation was needed to monitor reactor temperature and cooldown rate in the same mode. This issue was entered into the licensee's corrective action program under Condition Report 2005-02369. Failure to properly identify repairs needed to achieve and maintain cold shutdown following a control room fire and document them in analyses and procedures was a performance deficiency. This issue was more than minor because it affected the Mitigating Systems cornerstone attributes of protection from external factors (fire) and procedure quality. This finding was determined to have very low safety significance using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it involved an issue that only affected cold shutdown. Inspection Report# : 2005008(pdf) Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Control a High Radiation Area with Dose Rates Greater than One Rem per Hour The inspector reviewed a self-revealing non-cited violation of Technical Specification 5.7.3 involving the licensee's failure to control a high radiation area with dose rates greater than 1,000 millirem per hour. Specifically, on September 22, 2005, a radiation worker was performing a visual inspection of a low pressure coolant injection pipe penetration in the drywell. The worker climbed three feet above the floor elevation, at which time the worker's electronic dosimeter alarmed with peak dose rate of 582 millirem per hour. Radiation protection personnel performed a survey of the area and determined that dose rates were as high as 1,200 millirem per hour at one foot from the low pressure coolant injection pipe. This finding was entered into the licensee's corrective action program. This finding is greater than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involves the potential for a worker's unplanned or unintended dose resulting from actions contrary to technical specifications. When processed through the Occupational Radiation Safety Significance Determination Process, the finding is of very low safety significance because it did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2005004(pdf) Public Radiation Safety Physical Protection Physical Protection information not publicly available.

1Q/2006 Inspection Findings - Grand Gulf 1 Page 5 of 5 Miscellaneous Last modified : May 25, 2006

2Q/2006 Inspection Findings - Grand Gulf 1 Page 1 of 5 Grand Gulf 1 2Q/2006 Plant Inspection Findings Initiating Events Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment The inspectors identified two examples of a noncited violation of 10 CFR 50.65, Maintenance Rule, for failing to include maintenance that could increase the likelihood of an initiating event in the plant risk assessment. On February 2, 2006 and again on March 28, 2006, the licensees risk assessment did not include maintenance activities that increased the likelihood of a reactor scram. The licensee entered this into their corrective action program as Condition Reports CR-GGN-2006-1041 and CR-GGN-2006-1277. This finding is more than minor since the maintenance that was performed increased the likelihood of an initiating event. Using Inspection Manual Chapter 0609 Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is of very low safety significance since in both cases the change in incremental core damage probability and incremental large early release probability were less than 1E-6 and 1E-7, respectively. This finding has human performance crosscutting aspects because the inadequate risk assessments were due to personnel error. Inspection Report# : 2006002(pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: FIN Finding Plant Service Water Leak During Excavation The inspectors reviewed a self-revealing finding for a failure to follow procedure that resulted in a significant plant service water header leak. The licensee failed to review adequate documents to identify potential hazards as required by Procedure EN-S-112, Trenching, Excavation and Ground Penetrating Activities, Revision 2. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-0219. This finding is more than minor since it was associated with the human performance attribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of this finding has human performance cross-cutting aspects associated with a failure to follow procedures. Inspection Report# : 2006002(pdf) Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: FIN Finding Inadvertent Plant Service Water Pump Trip A Green self-revealing finding was identified for the inadvertent trip of a plant service water pump due to a failure to follow procedure. In addition, the procedure did not meet its stated purpose to verify the operation of a service water pump support system, specifically the well level indication system. The licensee entered this performance deficiency in their corrective action program for resolution. This finding is more than minor since it affected the configuration control and human performance attributes of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. This finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: FIN Finding Improper Maintenance Results in Partial Loss of Component Cooling Water A Green self-revealing finding was reviewed involving the failure of a newly installed corrosion monitor probe that resulted in a leak in the component cooling water system. Licensee personnel used an inadequate procedure to install the probe and therefore failed to verify the pressure retaining capability of the probe prior to installation. The licensee entered this performance deficiency in their corrective action program for resolution. This finding is more than minor since it affected the design control attribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the

2Q/2006 Inspection Findings - Grand Gulf 1 Page 2 of 5 finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Mitigating Systems Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Associated with Condensate Storage Tank Level Instrumentation The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to take prompt corrective actions to address a design deficiency involving condensate storage tank level instrumentation. The licensee identified the design deficiency on April 30, 1999, and issued compensatory actions for the operators to manually transfer high pressure core spray and reactor core isolation cooling from the condensate storage tank to the suppression pool in the event of failure of the tank. The licensee corrected the design deficiency on December 8, 2005. The licensee entered this issue in their corrective action program as CR-GGN-2006-1096. This finding is more than minor because it affected the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was of very low safety significance because it was a design deficiency that did not result in a loss of operability. This finding had cross-cutting aspects associated with problem identification and resolution in that station personnel did not implement corrective actions in a timely manner. Inspection Report# : 2006002(pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Alternate Shutdown Cooling Mode Not Properly Implemented in Alternate Shutdown Procedure. The inspectors identified a Green noncited violation for failure to have an alternative shutdown procedure to restore power following a control room evacuation with loss of offsite power that was independent of the control room. The licensee entered this into their corrective action program as CR-GGN-2005-1854. This finding is more than minor because it affected the mitigating systems cornerstone objective for the procedure quality and protection from external factors attributes. A Region IV Senior Reactor Analyst made a visit to the site during the week of January 30, 2006. Through discussions with engineers and walkdowns in the plant, the Senior Reactor Analyst determined that there is a credible fire scenario which could simultaneously cause a control room evacuation, a loss of offsite power, and prevent automatic starting and loading of the Division 1 emergency diesel generator. This issue was categorized as a postfire safe shutdown issue associated with response procedure quality. The degradation rating was determined to be Low because operator experience and familiarity with performing the required response actions were adequate to overcome the procedure deficiency. Therefore, this issue screened as having very low safety significance in Phase 1 of Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. Inspection Report# : 2006002(pdf) Significance: Mar 27, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a testing program to demonstrate the ability of standby service water-cooled heat exchangers The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to implement a testing program to demonstrate the ability of standby service water-cooled heat exchangers to perform their design basis functions under all conditions. The finding is greater than minor because, if left uncorrected, it would lead to a more significant issue, namely a heat exchanger would become unable to fulfill its safety function due to excessive fouling accumulating during the time between testing. This finding has cross-cutting aspects because it is more than minor, it represents current performance, and the cause is directly associated with the problem identification and resolution attribute of evaluation of test data Inspection Report# : 2006008(pdf) Significance: Mar 27, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate all design basis information into specifications and procedures

2Q/2006 Inspection Findings - Grand Gulf 1 Page 3 of 5 The team identified a finding of very low safety significance for a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to translate all design basis information into specifications and procedures were not adequate to assure that instrument uncertainties were correctly accounted for in the development of Technical Specification values or in the surveillance test acceptance criteria. The team determined this finding to be greater than minor because, similar to an example in MC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, the failure of licensee personnel to demonstrate where, and how, instrument uncertainties were translated into either Technical Specification values or the surveillance test acceptance criteria could result in systems and/or components not being capable of performing its design basis functions. This finding has cross-cutting aspects because it is more than minor, the failure to correct a previously identified adverse condition is an ongoing performance deficiency, and the cause (i.e., not understanding how to address instrument uncertainties) is directly associated with the problem identification and resolution attribute of corrective actions. Inspection Report# : 2006008(pdf) Significance: Oct 13, 2005 Identified By: NRC Item Type: NCV NonCited Violation Foreign Material in the Drywell The inspectors identified a Green noncited violation (NCV) of TS 5.4.1(a) for the failure of licensee personnel to perform an adequate drywell closeout inspection for foreign material. On October 13, 2005, licensee personnel failed to follow Integrated Operating Instruction 3-1-01-1, "Cold Shutdown to Minimum Generator Load," Attachment II, steps 2, 16, 21, and 34 of the drywell closeout sheet. The inspectors conducted a general inspection of the drywell and discovered approximately 50 foreign material items totaling a volume of approximately one and a half cubic feet in the drywell floor area. This foreign material included plastic wrappings and tie-wraps, articles of protective clothing, loose paper, metal objects and other miscellaneous material. This issue was entered into the licensee's corrective action program as CR-GGN-2006-00236. The finding is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding is determined to have very low safety significance because there was no loss of safety function of the emergency core cooling system suction strainers. The cause of the finding is related to the crosscutting element of human performance in that licensee personnel did not follow the drywell closeout procedure. Inspection Report# : 2005005(pdf) Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Disabling of Diesel Generator Alarms due to Failure to Follow Procedure The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1 involving a failure to follow procedure that resulted in the disabling of required supervisory alarms on the Division II emergency diesel generator. Specifically, operators failed to reset the alarm panel following routine testing. The licensee entered this performance deficiency into their corrective action program. This finding is more than minor since the disabling of required alarm functions for the emergency diesel generators could become a more significant safety concern if left uncorrected. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not result in an actual loss of the safety function. This finding also had crosscutting aspects associated with human performance. Inspection Report# : 2005004(pdf) Barrier Integrity Significance: Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Improper Reactor Recirculation Pump Speed Change The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for the failure to follow the procedure for reactor recirculation pump speed changes. Operators attempted to shift Recirculation Pump A to fast speed without verifying that interlocks were satisfied (annunciators not lit) as required by procedure. As a result, Recirculation Pump A failed to shift to fast speed, creating a flow mismatch between the recirculation loops. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-2329. This finding is more than minor since the failure to follow procedures regarding reactor manipulation, if left uncorrected, could lead to a more significant safety concern. The inspectors determined this finding affected the Barrier Integrity cornerstone since matched recirculation loop flows is an assumption used in the accident analysis for a loss-of-coolant accident resulting from a loop break. A flow mismatch could result in core response more severe than assumed in the accident analysis. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is of very low safety significance since it only affects the fuel barrier. This finding has crosscutting aspects associated with human performance since operators failed to follow procedures and verify that all annunciators associated with the recirculation loop pump temperatures were extinguished prior to shifting Recirculation Pump A to fast speed. Operators made incorrect assumptions regarding the meaning

2Q/2006 Inspection Findings - Grand Gulf 1 Page 4 of 5 of the lit annunciator and the impact that it would have on their ability to shift the pump to fast speed. Inspection Report# : 2006003(pdf) Emergency Preparedness Occupational Radiation Safety Significance: Sep 30, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Control a High Radiation Area with Dose Rates Greater than One Rem per Hour The inspector reviewed a self-revealing non-cited violation of Technical Specification 5.7.3 involving the licensee's failure to control a high radiation area with dose rates greater than 1,000 millirem per hour. Specifically, on September 22, 2005, a radiation worker was performing a visual inspection of a low pressure coolant injection pipe penetration in the drywell. The worker climbed three feet above the floor elevation, at which time the worker's electronic dosimeter alarmed with peak dose rate of 582 millirem per hour. Radiation protection personnel performed a survey of the area and determined that dose rates were as high as 1,200 millirem per hour at one foot from the low pressure coolant injection pipe. This finding was entered into the licensee's corrective action program. This finding is greater than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective to ensure the adequate protection of a worker's health and safety from exposure to radiation. The finding involves the potential for a worker's unplanned or unintended dose resulting from actions contrary to technical specifications. When processed through the Occupational Radiation Safety Significance Determination Process, the finding is of very low safety significance because it did not involve ALARA planning or work controls, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Inspection Report# : 2005004(pdf) Public Radiation Safety Significance: Sep 15, 2005 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to comply with certificate of compliance instructions The team reviewed the details associated with a self-revealing, non-cited violation of 10 CFR 71.17(c)(2) that resulted from the licensees failure to properly assemble a transportation package. The licensee was alerted to the error when a worker cleaning the assembly area discovered an unused reinforcing spacer block. However, the shipment had already been released from the site. The shipment was returned to the licensees facility approximately seven hours after it left. The licensee reported the occurrence in accordance with 10 CFR 71.95(c) and documented it in the corrective action program as CR-GGN-2005-01007. The finding is more than minor because it was associated with one of the Public Radiation Safety Cornerstone attributes (Transportation Program) and it affected the associated cornerstone objective in that the use of a shipping package not assembled in accordance with the certificate of compliance diminished the licensees ability to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain. The finding involved an occurrence in the licensees radioactive material transportation program that is contrary to NRC regulations; therefore, it was processed through the Public Radiation Safety Significance Determination Process. When the finding was processed through the significance determination process, it was found to have very low safety significance because: (1) it involved radioactive material control, (2) it was associated with transportation, (3) no radiation limit was exceeded, (4) there was no breach of the package during transit, (5) it was a certificate of compliance finding, (6) there was no design documentation deficiency, (7) it was not a maintenance/use performance deficiency, (8) it involved minor content deficiencies (minor structural component left out), but (9) it did not involve a major content deficiency. This finding also had cross-cutting aspects associated with human performance. This finding also had cross-cutting aspects associated with human performance, in that the failure of licensee personnel to comply with the certificate of compliance instructions directly resulted in the finding. Inspection Report# : 2005011(pdf) Physical Protection

2Q/2006 Inspection Findings - Grand Gulf 1 Page 5 of 5 Physical Protection information not publicly available. Miscellaneous Last modified : August 25, 2006

3Q/2006 Inspection Findings - Grand Gulf 1 Page 1 of 5 Grand Gulf 1 3Q/2006 Plant Inspection Findings Initiating Events Significance: Sep 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Document Deficiencies in the Corrective Action Program A finding was identified for failure to implement adequate controls to maintain the integrity of the 34.5 kV switchyard animal intrusion fence and for failure to initiate condition reports when the fence was found de-energized or the gate found open. The animal intrusion resulted in a reactor scram and an excessive reactor coolant system cooldown on February 11, 2005. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2006-3139. The finding was greater than minor because it affected the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The finding was determined to be of very low safety significance following completion of a modified Phase 2 significance determination process analysis. Although the NRC identified a performance deficiency related to maintaining the integrity of the animal intrusion fence and for failure to enter events into the corrective action program, the inspectors determined that no violation of regulatory requirements had occurred. In response to this event, the licensee revised operations procedures to require inspection of the switchyard fence conditions and required documenting deficiencies in their corrective action program. This item had cross cutting aspects related to human performance because procedures did not direct nonlicensed operators to monitor the condition of the fence. In addition, this item had crosscutting aspects related to problem identification and resolution because the licensee did not effectively implement corrective actions. Inspection Report# : 2006004(pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment The inspectors identified two examples of a noncited violation of 10 CFR 50.65, Maintenance Rule, for failing to include maintenance that could increase the likelihood of an initiating event in the plant risk assessment. On February 2, 2006 and again on March 28, 2006, the licensees risk assessment did not include maintenance activities that increased the likelihood of a reactor scram. The licensee entered this into their corrective action program as Condition Reports CR-GGN-2006-1041 and CR-GGN-2006-1277. This finding is more than minor since the maintenance that was performed increased the likelihood of an initiating event. Using Inspection Manual Chapter 0609 Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is of very low safety significance since in both cases the change in incremental core damage probability and incremental large early release probability were less than 1E-6 and 1E-7, respectively. This finding has human performance crosscutting aspects because the inadequate risk assessments were due to personnel error. Inspection Report# : 2006002(pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: FIN Finding Plant Service Water Leak During Excavation The inspectors reviewed a self-revealing finding for a failure to follow procedure that resulted in a significant plant service water header leak. The licensee failed to review adequate documents to identify potential hazards as required by Procedure EN-S-112, Trenching, Excavation and Ground Penetrating Activities, Revision 2. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-0219. This finding is more than minor since it was associated with the human performance attribute of the initiating events

3Q/2006 Inspection Findings - Grand Gulf 1 Page 2 of 5 cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of this finding has human performance cross-cutting aspects associated with a failure to follow procedures. Inspection Report# : 2006002(pdf) Mitigating Systems Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Loose Items in Safety Related Areas The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for a failure to control loose items in safety related structures. Between July 25 and September 13, 2006, the inspectors identified six examples of loose items in the auxiliary building and control building that did not meet the requirements of plant loose item control procedures. The licensee entered this issue in their corrective action program as CR-GGN-2006-3836. The failure to control loose items in the vicinity of safety related equipment was a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of protection against external factors (seismic) and affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability. The cause of this finding is related to the cross-cutting element of human performance in that licensee work practices did not effectively define and communicate expectations regarding compliance with plant procedures for the control of loose items in safety related structures. Inspection Report# : 2006004(pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Associated with Condensate Storage Tank Level Instrumentation The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to take prompt corrective actions to address a design deficiency involving condensate storage tank level instrumentation. The licensee identified the design deficiency on April 30, 1999, and issued compensatory actions for the operators to manually transfer high pressure core spray and reactor core isolation cooling from the condensate storage tank to the suppression pool in the event of failure of the tank. The licensee corrected the design deficiency on December 8, 2005. The licensee entered this issue in their corrective action program as CR-GGN-2006-1096. This finding is more than minor because it affected the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was of very low safety significance because it was a design deficiency that did not result in a loss of operability. This finding had cross-cutting aspects associated with problem identification and resolution in that station personnel did not implement corrective actions in a timely manner. Inspection Report# : 2006002(pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Alternate Shutdown Cooling Mode Not Properly Implemented in Alternate Shutdown Procedure. The inspectors identified a Green noncited violation for failure to have an alternative shutdown procedure to restore power following a control room evacuation with loss of offsite power that was independent of the control room. The licensee

3Q/2006 Inspection Findings - Grand Gulf 1 Page 3 of 5 entered this into their corrective action program as CR-GGN-2005-1854. This finding is more than minor because it affected the mitigating systems cornerstone objective for the procedure quality and protection from external factors attributes. A Region IV Senior Reactor Analyst made a visit to the site during the week of January 30, 2006. Through discussions with engineers and walkdowns in the plant, the Senior Reactor Analyst determined that there is a credible fire scenario which could simultaneously cause a control room evacuation, a loss of offsite power, and prevent automatic starting and loading of the Division 1 emergency diesel generator. This issue was categorized as a postfire safe shutdown issue associated with response procedure quality. The degradation rating was determined to be Low because operator experience and familiarity with performing the required response actions were adequate to overcome the procedure deficiency. Therefore, this issue screened as having very low safety significance in Phase 1 of Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. Inspection Report# : 2006002(pdf) Significance: Mar 27, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a testing program to demonstrate the ability of standby service water-cooled heat exchangers The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to implement a testing program to demonstrate the ability of standby service water-cooled heat exchangers to perform their design basis functions under all conditions. The finding is greater than minor because, if left uncorrected, it would lead to a more significant issue, namely a heat exchanger would become unable to fulfill its safety function due to excessive fouling accumulating during the time between testing. This finding has cross-cutting aspects because it is more than minor, it represents current performance, and the cause is directly associated with the problem identification and resolution attribute of evaluation of test data Inspection Report# : 2006008(pdf) Significance: Mar 27, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate all design basis information into specifications and procedures The team identified a finding of very low safety significance for a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to translate all design basis information into specifications and procedures were not adequate to assure that instrument uncertainties were correctly accounted for in the development of Technical Specification values or in the surveillance test acceptance criteria. The team determined this finding to be greater than minor because, similar to an example in MC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, the failure of licensee personnel to demonstrate where, and how, instrument uncertainties were translated into either Technical Specification values or the surveillance test acceptance criteria could result in systems and/or components not being capable of performing its design basis functions. This finding has cross-cutting aspects because it is more than minor, the failure to correct a previously identified adverse condition is an ongoing performance deficiency, and the cause (i.e., not understanding how to address instrument uncertainties) is directly associated with the problem identification and resolution attribute of corrective actions. Inspection Report# : 2006008(pdf) Significance: Oct 13, 2005 Identified By: NRC Item Type: NCV NonCited Violation Foreign Material in the Drywell The inspectors identified a Green noncited violation (NCV) of TS 5.4.1(a) for the failure of licensee personnel to perform an adequate drywell closeout inspection for foreign material. On October 13, 2005, licensee personnel failed to follow Integrated Operating Instruction 3-1-01-1, "Cold Shutdown to Minimum Generator Load," Attachment II, steps 2, 16, 21, and 34 of the drywell closeout sheet. The inspectors conducted a general inspection of the drywell and discovered

3Q/2006 Inspection Findings - Grand Gulf 1 Page 4 of 5 approximately 50 foreign material items totaling a volume of approximately one and a half cubic feet in the drywell floor area. This foreign material included plastic wrappings and tie-wraps, articles of protective clothing, loose paper, metal objects and other miscellaneous material. This issue was entered into the licensee's corrective action program as CR-GGN-2006-00236. The finding is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associated cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding is determined to have very low safety significance because there was no loss of safety function of the emergency core cooling system suction strainers. The cause of the finding is related to the crosscutting element of human performance in that licensee personnel did not follow the drywell closeout procedure. Inspection Report# : 2005005(pdf) Barrier Integrity Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Leakage Detection Sensing Lines The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to maintain the original design configuration of a leakage detection instrument sensing line in the fuel pool cooling and cleanup system. The licensee entered this issue in their corrective action program as CR-GGN-2006-3569. This finding is more than minor since it affects the design control attribute of the spent fuel pool cooling aspect of the Barrier Integrity cornerstone and affects the cornerstone objective of providing assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it only affected the radiological barrier function provided by the spent fuel pool. Inspection Report# : 2006004(pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Containment Pool Liner Leakage The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure of the licensee to take actions required by operator rounds in response to containment pool liner leakage. The licensee entered this issue into their corrective action program as CR-GGN-2006-3500. The finding was more than minor since the failure of operators to perform operator rounds could lead to a more significant safety concern if left uncorrected. Additionally, the identified liner leakage represented a degrading condition that, if left uncorrected, could continue to degrade and could potentially result in the migration of water to other portions of the containment structure. The inspectors determined this finding affected the Barrier Integrity cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it does not represent an actual open pathway in the physical integrity of the reactor containment or an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. The cause of this finding is related to the crosscutting element of human performance in that licensee work practices did not effectively define and communicate expectations regarding compliance with plant procedures for the conduct of operator rounds. Inspection Report# : 2006004(pdf) Significance: Jun 30, 2006 Identified By: NRC

3Q/2006 Inspection Findings - Grand Gulf 1 Page 5 of 5 Item Type: NCV NonCited Violation Improper Reactor Recirculation Pump Speed Change The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for the failure to follow the procedure for reactor recirculation pump speed changes. Operators attempted to shift Recirculation Pump A to fast speed without verifying that interlocks were satisfied (annunciators not lit) as required by procedure. As a result, Recirculation Pump A failed to shift to fast speed, creating a flow mismatch between the recirculation loops. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-2329. This finding is more than minor since the failure to follow procedures regarding reactor manipulation, if left uncorrected, could lead to a more significant safety concern. The inspectors determined this finding affected the Barrier Integrity cornerstone since matched recirculation loop flows is an assumption used in the accident analysis for a loss-of-coolant accident resulting from a loop break. A flow mismatch could result in core response more severe than assumed in the accident analysis. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is of very low safety significance since it only affects the fuel barrier. This finding has crosscutting aspects associated with human performance since operators failed to follow procedures and verify that all annunciators associated with the recirculation loop pump temperatures were extinguished prior to shifting Recirculation Pump A to fast speed. Operators made incorrect assumptions regarding the meaning of the lit annunciator and the impact that it would have on their ability to shift the pump to fast speed. Inspection Report# : 2006003(pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available. Miscellaneous Last modified : December 21, 2006

4Q/2006 Inspection Findings - Grand Gulf 1 Page 1 of 6 Grand Gulf 1 4Q/2006 Plant Inspection Findings Initiating Events Significance: Dec 31, 2006 Identified By: NRC Item Type: FIN Finding Insufficient Preventive Maintenance of Bus Duct Cooling System Results in Unplanned Power Reduction The inspectors reviewed a Green, self-revealing finding for failure to implement preventive maintenance on the bus duct cooling system components prior to system failures, causing a plant transient. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-3996. The finding is more than minor since it affects the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding has a very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood (Section 4OA3). Inspection Report# : 2006005 (pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Document Deficiencies in the Corrective Action Program A finding was identified for failure to implement adequate controls to maintain the integrity of the 34.5 kV switchyard animal intrusion fence and for failure to initiate condition reports when the fence was found de-energized or the gate found open. The animal intrusion resulted in a reactor scram and an excessive reactor coolant system cooldown on February 11, 2005. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2006-3139. The finding was greater than minor because it affected the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The finding was determined to be of very low safety significance following completion of a modified Phase 2 significance determination process analysis. Although the NRC identified a performance deficiency related to maintaining the integrity of the animal intrusion fence and for failure to enter events into the corrective action program, the inspectors determined that no violation of regulatory requirements had occurred. In response to this event, the licensee revised operations procedures to require inspection of the switchyard fence conditions and required documenting deficiencies in their corrective action program. This item had cross cutting aspects related to human performance because procedures did not direct nonlicensed operators to monitor the condition of the fence. In addition, this item had crosscutting aspects related to problem identification and resolution because the licensee did not effectively implement corrective actions. Inspection Report# : 2006004 (pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment The inspectors identified two examples of a noncited violation of 10 CFR 50.65, Maintenance Rule, for failing to include maintenance that could increase the likelihood of an initiating event in the plant risk assessment. On February 2, 2006 and again on March 28, 2006, the licensees risk assessment did not include maintenance activities that increased the likelihood of a reactor scram. The licensee entered this into their corrective action program as Condition Reports CR-GGN-2006-1041 and CR-GGN-2006-1277. This finding is more than minor since the maintenance that was performed increased the likelihood of an initiating event. Using Inspection Manual Chapter 0609 Appendix K, Maintenance Risk Assessment and Risk Management Significance

4Q/2006 Inspection Findings - Grand Gulf 1 Page 2 of 6 Determination Process, the finding is of very low safety significance since in both cases the change in incremental core damage probability and incremental large early release probability were less than 1E-6 and 1E-7, respectively. This finding has human performance crosscutting aspects because the inadequate risk assessments were due to personnel error. Inspection Report# : 2006002 (pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: FIN Finding Plant Service Water Leak During Excavation The inspectors reviewed a self-revealing finding for a failure to follow procedure that resulted in a significant plant service water header leak. The licensee failed to review adequate documents to identify potential hazards as required by Procedure EN-S-112, Trenching, Excavation and Ground Penetrating Activities, Revision 2. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-0219. This finding is more than minor since it was associated with the human performance attribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of this finding has human performance cross-cutting aspects associated with a failure to follow procedures. Inspection Report# : 2006002 (pdf) Mitigating Systems Significance: Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Station Procedures for Conducting Maintenance Activities The inspectors reviewed a Green, self-revealing noncited violation of Technical Specification 5.4.1(a) for failure to follow station maintenance procedures while troubleshooting the control rod drive Pump A hand switch green indicating light socket. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-4474. The finding is more than minor since it affects the human performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, improper maintenance practices on control room equipment could lead to a more significant safety concern. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, inspectors determined that the finding has very low safety significance because it did not result in a loss of safety function. This finding has a crosscutting aspect in the area of human performance associated with work practices in that licensee personnel proceeded to troubleshoot the bulb in the face of uncertainty surrounding the required bulb type and expected system response (Section 1R19). Inspection Report# : 2006005 (pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Loose Items in Safety Related Areas The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for a failure to control loose items in safety related structures. Between July 25 and September 13, 2006, the inspectors identified six examples of loose items in the auxiliary building and control building that did not meet the requirements of plant loose item control procedures. The licensee entered this issue in their corrective action program as CR-GGN-2006-3836. The failure to control loose items in the vicinity of safety related equipment was a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of protection against external

4Q/2006 Inspection Findings - Grand Gulf 1 Page 3 of 6 factors (seismic) and affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability. The cause of this finding is related to the cross-cutting element of human performance in that licensee work practices did not effectively define and communicate expectations regarding compliance with plant procedures for the control of loose items in safety related structures. Inspection Report# : 2006004 (pdf) Significance: Aug 07, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Standby Diesel Generator Cylinder Head Failures A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to preclude repetition of a significant condition adverse to quality. Specifically, the licensee failed to take actions to prevent subsequent standby diesel generator engine head failures attributed to corrosion fatigue in 1992, 1996, and 2006. This issue was entered into the licensee's corrective action program as Conditon Report CR-GGN-2006-1955. The finding was more than minor since it affected the Mitigation System Cornerstone attribute of availability and reliability of mitigating equipment, specifically the standby diesel generators. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding is of very low significance since it only involved the loss of one train of diesel generators for less than the technical specification allowed outage time (Section 4.b). Inspection Report# : 2006010 (pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Associated with Condensate Storage Tank Level Instrumentation The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to take prompt corrective actions to address a design deficiency involving condensate storage tank level instrumentation. The licensee identified the design deficiency on April 30, 1999, and issued compensatory actions for the operators to manually transfer high pressure core spray and reactor core isolation cooling from the condensate storage tank to the suppression pool in the event of failure of the tank. The licensee corrected the design deficiency on December 8, 2005. The licensee entered this issue in their corrective action program as CR-GGN-2006-1096. This finding is more than minor because it affected the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was of very low safety significance because it was a design deficiency that did not result in a loss of operability. This finding had cross-cutting aspects associated with problem identification and resolution in that station personnel did not implement corrective actions in a timely manner. Inspection Report# : 2006002 (pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Alternate Shutdown Cooling Mode Not Properly Implemented in Alternate Shutdown Procedure. The inspectors identified a Green noncited violation for failure to have an alternative shutdown procedure to restore power following a control room evacuation with loss of offsite power that was independent of the control room. The licensee entered this into their corrective action program as CR-GGN-2005-1854. This finding is more than minor because it affected the mitigating systems cornerstone objective for the procedure quality and protection from external factors attributes. A Region IV Senior Reactor Analyst made a visit to the site during the week of January 30, 2006. Through discussions with engineers and walkdowns in the plant, the Senior Reactor Analyst determined that there is a credible fire scenario which could simultaneously cause a control room evacuation, a loss of offsite power, and prevent automatic starting and loading of the Division 1 emergency diesel generator. This issue was categorized as a postfire safe shutdown issue associated with response procedure quality. The degradation rating was determined to be Low because operator experience and familiarity with performing the required response actions were

4Q/2006 Inspection Findings - Grand Gulf 1 Page 4 of 6 adequate to overcome the procedure deficiency. Therefore, this issue screened as having very low safety significance in Phase 1 of Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. Inspection Report# : 2006002 (pdf) Significance: Mar 27, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement a testing program to demonstrate the ability of standby service water-cooled heat exchangers The team identified a finding of very low safety significance involving a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to implement a testing program to demonstrate the ability of standby service water-cooled heat exchangers to perform their design basis functions under all conditions. The finding is greater than minor because, if left uncorrected, it would lead to a more significant issue, namely a heat exchanger would become unable to fulfill its safety function due to excessive fouling accumulating during the time between testing. This finding has cross-cutting aspects because it is more than minor, it represents current performance, and the cause is directly associated with the problem identification and resolution attribute of evaluation of test data Inspection Report# : 2006008 (pdf) Significance: Mar 27, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to translate all design basis information into specifications and procedures The team identified a finding of very low safety significance for a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to translate all design basis information into specifications and procedures were not adequate to assure that instrument uncertainties were correctly accounted for in the development of Technical Specification values or in the surveillance test acceptance criteria. The team determined this finding to be greater than minor because, similar to an example in MC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, the failure of licensee personnel to demonstrate where, and how, instrument uncertainties were translated into either Technical Specification values or the surveillance test acceptance criteria could result in systems and/or components not being capable of performing its design basis functions. This finding has cross-cutting aspects because it is more than minor, the failure to correct a previously identified adverse condition is an ongoing performance deficiency, and the cause (i.e., not understanding how to address instrument uncertainties) is directly associated with the problem identification and resolution attribute of corrective actions. Inspection Report# : 2006008 (pdf) Barrier Integrity Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Leakage Detection Sensing Lines The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to maintain the original design configuration of a leakage detection instrument sensing line in the fuel pool cooling and cleanup system. The licensee entered this issue in their corrective action program as CR-GGN-2006-3569. This finding is more than minor since it affects the design control attribute of the spent fuel pool cooling aspect of the Barrier Integrity cornerstone and affects the cornerstone objective of providing assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it only affected the

4Q/2006 Inspection Findings - Grand Gulf 1 Page 5 of 6 radiological barrier function provided by the spent fuel pool. Inspection Report# : 2006004 (pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Containment Pool Liner Leakage The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure of the licensee to take actions required by operator rounds in response to containment pool liner leakage. The licensee entered this issue into their corrective action program as CR-GGN-2006-3500. The finding was more than minor since the failure of operators to perform operator rounds could lead to a more significant safety concern if left uncorrected. Additionally, the identified liner leakage represented a degrading condition that, if left uncorrected, could continue to degrade and could potentially result in the migration of water to other portions of the containment structure. The inspectors determined this finding affected the Barrier Integrity cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it does not represent an actual open pathway in the physical integrity of the reactor containment or an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. The cause of this finding is related to the crosscutting element of human performance in that licensee work practices did not effectively define and communicate expectations regarding compliance with plant procedures for the conduct of operator rounds. Inspection Report# : 2006004 (pdf) Significance: Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Improper Reactor Recirculation Pump Speed Change The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for the failure to follow the procedure for reactor recirculation pump speed changes. Operators attempted to shift Recirculation Pump A to fast speed without verifying that interlocks were satisfied (annunciators not lit) as required by procedure. As a result, Recirculation Pump A failed to shift to fast speed, creating a flow mismatch between the recirculation loops. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-2329. This finding is more than minor since the failure to follow procedures regarding reactor manipulation, if left uncorrected, could lead to a more significant safety concern. The inspectors determined this finding affected the Barrier Integrity cornerstone since matched recirculation loop flows is an assumption used in the accident analysis for a loss-of-coolant accident resulting from a loop break. A flow mismatch could result in core response more severe than assumed in the accident analysis. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is of very low safety significance since it only affects the fuel barrier. This finding has crosscutting aspects associated with human performance since operators failed to follow procedures and verify that all annunciators associated with the recirculation loop pump temperatures were extinguished prior to shifting Recirculation Pump A to fast speed. Operators made incorrect assumptions regarding the meaning of the lit annunciator and the impact that it would have on their ability to shift the pump to fast speed. Inspection Report# : 2006003 (pdf) Emergency Preparedness Occupational Radiation Safety

4Q/2006 Inspection Findings - Grand Gulf 1 Page 6 of 6 Public Radiation Safety Physical Protection Physical Protection information not publicly available. Miscellaneous Last modified : March 01, 2007

Grand Gulf 1 1Q/2007 Plant Inspection Findings Initiating Events Significance: Dec 31, 2006 Identified By: NRC Item Type: FIN Finding Insufficient Preventive Maintenance of Bus Duct Cooling System Results in Unplanned Power Reduction The inspectors reviewed a Green, self-revealing finding for failure to implement preventive maintenance on the bus duct cooling system components prior to system failures, causing a plant transient. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-3996. The finding is more than minor since it affects the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding has a very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood (Section 4OA3). Inspection Report# : 2006005 (pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Document Deficiencies in the Corrective Action Program A finding was identified for failure to implement adequate controls to maintain the integrity of the 34.5 kV switchyard animal intrusion fence and for failure to initiate condition reports when the fence was found de-energized or the gate found open. The animal intrusion resulted in a reactor scram and an excessive reactor coolant system cooldown on February 11, 2005. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2006-3139. The finding was greater than minor because it affected the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The finding was determined to be of very low safety significance following completion of a modified Phase 2 significance determination process analysis. Although the NRC identified a performance deficiency related to maintaining the integrity of the animal intrusion fence and for failure to enter events into the corrective action program, the inspectors determined that no violation of regulatory requirements had occurred. In response to this event, the licensee revised operations procedures to require inspection of the switchyard fence conditions and required documenting deficiencies in their corrective action program. This item had cross cutting aspects related to human performance because procedures did not direct nonlicensed operators to monitor the condition of the fence. In addition, this item had crosscutting aspects related to problem identification and resolution because the licensee did not effectively implement corrective actions. Inspection Report# : 2006004 (pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment The inspectors identified two examples of a noncited violation of 10 CFR 50.65, Maintenance Rule, for failing to include maintenance that could increase the likelihood of an initiating event in the plant risk assessment. On February 2, 2006 and again on March 28, 2006, the licensees risk assessment did not include maintenance activities that increased the likelihood of a reactor scram. The licensee entered this into their corrective action program as Condition Reports CR-GGN-2006-1041 and CR-GGN-2006-1277. This finding is more than minor since the maintenance that was performed increased the likelihood of an initiating event. Using Inspection Manual Chapter 0609 Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the finding is of very low safety significance since in both cases the change in incremental core

damage probability and incremental large early release probability were less than 1E-6 and 1E-7, respectively. This finding has human performance crosscutting aspects because the inadequate risk assessments were due to personnel error. Inspection Report# : 2006002 (pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: FIN Finding Plant Service Water Leak During Excavation The inspectors reviewed a self-revealing finding for a failure to follow procedure that resulted in a significant plant service water header leak. The licensee failed to review adequate documents to identify potential hazards as required by Procedure EN-S-112, Trenching, Excavation and Ground Penetrating Activities, Revision 2. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-0219. This finding is more than minor since it was associated with the human performance attribute of the initiating events cornerstone and directly affected the cornerstone objective of limiting events that challenge plant stability. Based on the results of a Significance Determination Process Phase 1 evaluation, the finding is of very low safety significance (Green) since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood. The cause of this finding has human performance cross-cutting aspects associated with a failure to follow procedures. Inspection Report# : 2006002 (pdf) Mitigating Systems Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation. The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving the failure to follow procedures resulted in an inadequate operability evaluation for a degraded switchgear ventilation system. Specifically, the evaluation utilized several non-conservative input assumptions and failed to adequately evaluate the potential adverse affects from changing weather conditions. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-0554. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it did not result in a loss of operability. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions (Section 1R15). Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Command and Control Results in Inappropriate Valve Manipulations. The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for failure to meet procedural requirements involving command and control in the control room. Specifically, the control room supervisor was not informed of a system alignment change directed by the shift technical advisor. The licensee entered this issue in their corrective action program as CR-GGN-2007-1060. This finding is more than minor since the failure to maintain appropriate command and control in the control room, if left uncorrected, could lead to a more significant safety concern. The inspectors determined that this finding affected the mitigating systems cornerstone. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, the finding is of very low safety significance since it did not result in an actual loss of operability. This finding has a crosscutting aspect in the area of human performance associated with work practices because the failure to communicate

the system realignment to the control room supervisor prevented the control room supervisor from maintaining proper supervisory oversight of work activities. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Resulting in Isolation of Switchgear Room Ventilation. A self-revealing Green noncited violation of Technical Specification 5.4.1(a) was identified for the failure to follow a surveillance procedure resulting in the inadvertent isolation of ventilation to the Division 1 and Division 3 safety-related switchgear rooms. The licensee entered this issue in their corrective action program as CR-GGN-2006-4394. This finding is more than minor since it affected the human performance attribute of the mitigating systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. This finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel did not effectively utilize human error prevention techniques, such as self and peer checking. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Standby Service Water System Leakage. The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee failed to take adequate corrective actions in response to service water leakage from drywell purge compressor oil cooler drain plugs. The licensee entered this issue in their corrective action program as CR-GGN-2006-4762. This finding is more than minor because if left uncorrected, the zinc drain plugs could have deteriorated to a point at which service water leakage would have impacted the performance of the standby service water system. This finding also affects the equipment performance attribute of the mitigating systems cornerstone and impacts the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Using the Significance Determination Process Phase 1 Screening Worksheet in Appendix A of Inspection Manual Chapter 0609, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to thoroughly evaluate the cause and extent of condition for corrosion identified on the drain plugs of the Train B purge compressor oil cooler. Inspection Report# : 2007002 (pdf) Significance: Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Station Procedures for Conducting Maintenance Activities The inspectors reviewed a Green, self-revealing noncited violation of Technical Specification 5.4.1(a) for failure to follow station maintenance procedures while troubleshooting the control rod drive Pump A hand switch green indicating light socket. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-4474. The finding is more than minor since it affects the human performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, improper maintenance practices on control room equipment could lead to a more significant safety concern. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, inspectors determined that the finding has very low safety significance because it did not result in a loss of safety function. This finding has a crosscutting aspect in the area of human performance associated with work practices in that licensee personnel proceeded to troubleshoot the bulb in the face of uncertainty surrounding the required bulb type and expected system response (Section 1R19). Inspection Report# : 2006005 (pdf)

Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Loose Items in Safety Related Areas The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for a failure to control loose items in safety related structures. Between July 25 and September 13, 2006, the inspectors identified six examples of loose items in the auxiliary building and control building that did not meet the requirements of plant loose item control procedures. The licensee entered this issue in their corrective action program as CR-GGN-2006-3836. The failure to control loose items in the vicinity of safety related equipment was a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of protection against external factors (seismic) and affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability. The cause of this finding is related to the cross-cutting element of human performance in that licensee work practices did not effectively define and communicate expectations regarding compliance with plant procedures for the control of loose items in safety related structures. Inspection Report# : 2006004 (pdf) Significance: Aug 07, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Standby Diesel Generator Cylinder Head Failures A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to preclude repetition of a significant condition adverse to quality. Specifically, the licensee failed to take actions to prevent subsequent standby diesel generator engine head failures attributed to corrosion fatigue in 1992, 1996, and 2006. This issue was entered into the licensee's corrective action program as Conditon Report CR-GGN-2006-1955. The finding was more than minor since it affected the Mitigation System Cornerstone attribute of availability and reliability of mitigating equipment, specifically the standby diesel generators. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding is of very low significance since it only involved the loss of one train of diesel generators for less than the technical specification allowed outage time (Section 4.b). Inspection Report# : 2006010 (pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Associated with Condensate Storage Tank Level Instrumentation The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to take prompt corrective actions to address a design deficiency involving condensate storage tank level instrumentation. The licensee identified the design deficiency on April 30, 1999, and issued compensatory actions for the operators to manually transfer high pressure core spray and reactor core isolation cooling from the condensate storage tank to the suppression pool in the event of failure of the tank. The licensee corrected the design deficiency on December 8, 2005. The licensee entered this issue in their corrective action program as CR-GGN-2006-1096. This finding is more than minor because it affected the design control attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events. The finding was of very low safety significance because it was a design deficiency that did not result in a loss of operability. This finding had cross-cutting aspects associated with problem identification and resolution in that station personnel did not implement corrective actions in a timely manner. Inspection Report# : 2006002 (pdf) Significance: Apr 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation

Alternate Shutdown Cooling Mode Not Properly Implemented in Alternate Shutdown Procedure. The inspectors identified a Green noncited violation for failure to have an alternative shutdown procedure to restore power following a control room evacuation with loss of offsite power that was independent of the control room. The licensee entered this into their corrective action program as CR-GGN-2005-1854. This finding is more than minor because it affected the mitigating systems cornerstone objective for the procedure quality and protection from external factors attributes. A Region IV Senior Reactor Analyst made a visit to the site during the week of January 30, 2006. Through discussions with engineers and walkdowns in the plant, the Senior Reactor Analyst determined that there is a credible fire scenario which could simultaneously cause a control room evacuation, a loss of offsite power, and prevent automatic starting and loading of the Division 1 emergency diesel generator. This issue was categorized as a postfire safe shutdown issue associated with response procedure quality. The degradation rating was determined to be Low because operator experience and familiarity with performing the required response actions were adequate to overcome the procedure deficiency. Therefore, this issue screened as having very low safety significance in Phase 1 of Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. Inspection Report# : 2006002 (pdf) Barrier Integrity Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Leakage Detection Sensing Lines The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to maintain the original design configuration of a leakage detection instrument sensing line in the fuel pool cooling and cleanup system. The licensee entered this issue in their corrective action program as CR-GGN-2006-3569. This finding is more than minor since it affects the design control attribute of the spent fuel pool cooling aspect of the Barrier Integrity cornerstone and affects the cornerstone objective of providing assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it only affected the radiological barrier function provided by the spent fuel pool. Inspection Report# : 2006004 (pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Containment Pool Liner Leakage The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure of the licensee to take actions required by operator rounds in response to containment pool liner leakage. The licensee entered this issue into their corrective action program as CR-GGN-2006-3500. The finding was more than minor since the failure of operators to perform operator rounds could lead to a more significant safety concern if left uncorrected. Additionally, the identified liner leakage represented a degrading condition that, if left uncorrected, could continue to degrade and could potentially result in the migration of water to other portions of the containment structure. The inspectors determined this finding affected the Barrier Integrity cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it does not represent an actual open pathway in the physical integrity of the reactor containment or an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. The cause of this finding is related to the crosscutting element of human performance in that licensee work practices did not effectively define and communicate expectations regarding compliance with plant procedures for the conduct of operator rounds. Inspection Report# : 2006004 (pdf) Significance: Jun 30, 2006 Identified By: NRC

Item Type: NCV NonCited Violation Improper Reactor Recirculation Pump Speed Change The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for the failure to follow the procedure for reactor recirculation pump speed changes. Operators attempted to shift Recirculation Pump A to fast speed without verifying that interlocks were satisfied (annunciators not lit) as required by procedure. As a result, Recirculation Pump A failed to shift to fast speed, creating a flow mismatch between the recirculation loops. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-2329. This finding is more than minor since the failure to follow procedures regarding reactor manipulation, if left uncorrected, could lead to a more significant safety concern. The inspectors determined this finding affected the Barrier Integrity cornerstone since matched recirculation loop flows is an assumption used in the accident analysis for a loss-of-coolant accident resulting from a loop break. A flow mismatch could result in core response more severe than assumed in the accident analysis. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is of very low safety significance since it only affects the fuel barrier. This finding has crosscutting aspects associated with human performance since operators failed to follow procedures and verify that all annunciators associated with the recirculation loop pump temperatures were extinguished prior to shifting Recirculation Pump A to fast speed. Operators made incorrect assumptions regarding the meaning of the lit annunciator and the impact that it would have on their ability to shift the pump to fast speed. Inspection Report# : 2006003 (pdf) Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedural Guidance and Radiation Work Instructions While Supporting Radiograpy Operations A self-revealing, Green noncited violation of Technical Specification 5.4.1 was identified for the failure to follow procedural guidance and radiation work instructions while supporting radiography operations. All entrances to the area in which radiography was conducted were not barricaded and posted at the two millirem per hour point, as required. However, the high radiation area was barricaded, posted, and guarded. As immediate corrective action, the licensee postponed additional radiography and initiated a review of the occurrence. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective, in that the failure to control access to areas in which radiography is conducted could result in unplanned personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety signficance because (1) it was not an ALARA finding, (2) there was no overexposeure, (3) there was no substantial potential for an overexposure because no one entered the area in which high doses were possible, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work control because the licensee did not coordinate work activities by incorporating actions to address the need to keep personnel apprised of work status. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit Instructions Prohibiting High Radiation Area Entry. A self-revealing Green noncited violation of Technical Specification 5.4.1 was identified for the failure to follow radiation work permit instructions prohibiting high radiation area entry. Two outage workers entered a high radiation area on the 139-foot elevation of the auxiliary steam tunnel, in violation of their radiation work permit instructions. The licensee was alerted to the entry into the high radiation area by one of the workers alarming dosimeter. As immediate corrective action, the licensee revoked the worker's access to the radiologically controlled area. Further corrective action is being evaluated.

This finding is greater than minor because it is associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective, in that the failure to follow radiation work permit instructions could result in unplanned personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety significance because (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for overexposure because, at the highest dose rate, it would have taken 40 hours to receive a whole-body overexposure, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work practices because the workers failed to use error prevention techniques such as self and peer checking. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Radiological Hazard Caused by Water Leaking in the Drywell. The inspector identified a Green noncited violation of 10 CFR 20.1501(a) because the licensee failed to adequately evaluate the radiological hazard caused by water leaking from a valve in the drywell. The licensee failed to maintain knowledge of changing radiological conditions. As immediate corrective action, the licensee surveyed the area to obtain current information. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety significance because (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions in deciding the correct contamination survey frequency in the drywell. Inspection Report# : 2007002 (pdf) Public Radiation Safety Physical Protection Physical Protection information not publicly available. Miscellaneous Last modified : June 01, 2007

Grand Gulf 1 2Q/2007 Plant Inspection Findings Initiating Events Significance: Dec 31, 2006 Identified By: NRC Item Type: FIN Finding Insufficient Preventive Maintenance of Bus Duct Cooling System Results in Unplanned Power Reduction The inspectors reviewed a Green, self-revealing finding for failure to implement preventive maintenance on the bus duct cooling system components prior to system failures, causing a plant transient. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-3996. The finding is more than minor since it affects the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding has a very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood (Section 4OA3). Inspection Report# : 2006005 (pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Document Deficiencies in the Corrective Action Program A finding was identified for failure to implement adequate controls to maintain the integrity of the 34.5 kV switchyard animal intrusion fence and for failure to initiate condition reports when the fence was found de-energized or the gate found open. The animal intrusion resulted in a reactor scram and an excessive reactor coolant system cooldown on February 11, 2005. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2006-3139. The finding was greater than minor because it affected the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The finding was determined to be of very low safety significance following completion of a modified Phase 2 significance determination process analysis. Although the NRC identified a performance deficiency related to maintaining the integrity of the animal intrusion fence and for failure to enter events into the corrective action program, the inspectors determined that no violation of regulatory requirements had occurred. In response to this event, the licensee revised operations procedures to require inspection of the switchyard fence conditions and required documenting deficiencies in their corrective action program. This item had cross cutting aspects related to human performance because procedures did not direct nonlicensed operators to monitor the condition of the fence. In addition, this item had crosscutting aspects related to problem identification and resolution because the licensee did not effectively implement corrective actions. Inspection Report# : 2006004 (pdf) Mitigating Systems Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation. The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving the failure to follow procedures resulted in an inadequate operability evaluation for a degraded switchgear ventilation system.

Specifically, the evaluation utilized several non-conservative input assumptions and failed to adequately evaluate the potential adverse affects from changing weather conditions. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-0554. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it did not result in a loss of operability. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions (Section 1R15). Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Command and Control Results in Inappropriate Valve Manipulations. The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for failure to meet procedural requirements involving command and control in the control room. Specifically, the control room supervisor was not informed of a system alignment change directed by the shift technical advisor. The licensee entered this issue in their corrective action program as CR-GGN-2007-1060. This finding is more than minor since the failure to maintain appropriate command and control in the control room, if left uncorrected, could lead to a more significant safety concern. The inspectors determined that this finding affected the mitigating systems cornerstone. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, the finding is of very low safety significance since it did not result in an actual loss of operability. This finding has a crosscutting aspect in the area of human performance associated with work practices because the failure to communicate the system realignment to the control room supervisor prevented the control room supervisor from maintaining proper supervisory oversight of work activities. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Resulting in Isolation of Switchgear Room Ventilation. A self-revealing Green noncited violation of Technical Specification 5.4.1(a) was identified for the failure to follow a surveillance procedure resulting in the inadvertent isolation of ventilation to the Division 1 and Division 3 safety-related switchgear rooms. The licensee entered this issue in their corrective action program as CR-GGN-2006-4394. This finding is more than minor since it affected the human performance attribute of the mitigating systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. This finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel did not effectively utilize human error prevention techniques, such as self and peer checking. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Standby Service Water System Leakage. The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee failed to take adequate corrective actions in response to service water leakage from drywell purge compressor oil cooler drain plugs. The licensee entered this issue in their corrective action program as CR-GGN-2006-4762. This finding is more than minor because if left uncorrected, the zinc drain plugs could have deteriorated to a point at which service water leakage would have impacted the performance of the standby service water system. This finding also affects the equipment performance attribute of the mitigating systems cornerstone and impacts the cornerstone

objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Using the Significance Determination Process Phase 1 Screening Worksheet in Appendix A of Inspection Manual Chapter 0609, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to thoroughly evaluate the cause and extent of condition for corrosion identified on the drain plugs of the Train B purge compressor oil cooler. Inspection Report# : 2007002 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of High Standby Diesel Generator Temperatures (Section 3.0). The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to identify and correct the cause of elevated temperatures adversely affecting the safety function of the Division I standby diesel generator that had previously occurred in 1999 and 2004. Subsequently, on January 30, 2007, the Division I standby diesel generator again experienced elevated temperatures during a surveillance run and was subsequently declared inoperable. This issue was entered into the licensee's corrective action program as Condition Report GGN-2007-0378. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The Phase 1 Worksheets in Manual Chapter 0609, Significance Determination Process, were used to conclude that a Phase 2 analysis was required because the condition represented a loss of safety function of a single train of a Technical Specification system for greater than its allowed outage time. The inspectors performed a Phase 2 analysis using Appendix A, Technical Basis For At Power Significance Determination Process, of Manual Chapter 0609, Significance Determination Process, and the Phase 2 Worksheet for Grand Gulf. The Phase 2 evaluation concluded that the finding was of very low safety significance. A Phase 3 significance determination analysis also determined the finding was of very low safety significance. The cause of the finding is related to the problem identification and resolution crosscutting area in that the licensee failed to thoroughly evaluate the problem resulting in ineffective corrective actions being implemented that failed to prevent recurrence of a significant condition adverse to quality (Section 3.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Alarm Response Instruction for SDG High Jacket Water Temperature (Section 4.0). The team identified a noncited violation of Technical Specification 5.4.1 (a) involving the failure to maintain an adequate alarm response instruction for standby diesel generator high jacket water temperature. Specifically, Procedure 04-1-02-1H22-P400, Alarm Response Instruction, Panel No.: 1H-22-P400, Safety Related, Revision 109, failed to provide adequate guidance to manually override the standby diesel generator jacket water cooling system temperature control valve during emergency conditions. This issue was entered into the licensee's corrective action program as Condition Report GG-2007-1837. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of procedure quality and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance because it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating events. The cause of the finding is related to the problem identification and resolution crosscutting area in that the licensee did not take appropriate corrective actions to adequately address a previously identified safety concern (Section 4.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify a Degraded Condition (Section 5.0).

The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to promptly identify a condition adverse to quality. Between February 2-15, 2007, the licensee failed to promptly identify that corrective actions taken in response to a January 30, 2007, failure of the Division 1 standby diesel generator jacket water cooling system temperature control valve had not addressed the cause of the valve failure. Specifically, following the valves failure, the licensee inappropriately concluded the valves internal thermal elements were faulty, replaced the elements, performed postmaintenance testing, and declared the valve and associated standby diesel generator operable on February 1, 2007. Subsequent testing of the suspect faulty thermal elements on February 2 and 13, 2007, found the components were functional. Following receipt of the testing results, the licensee failed to promptly identify that replacement of the thermal elements failed to address the cause of the problem resulting in the failure to evaluate a potential degraded condition affecting operability of the standby emergency diesel generator. This issue was entered into the licensee's corrective action program as Condition Report GGN-2007-2255. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associate cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance because the condition did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating events. The cause of the finding is related to the problem identification and resolution crosscutting area in that the licensee did not identify an issue completely, accurately, and in a timely manner commensurate with its safety significance resulting in the failure to evaluate a potential degraded condition for operability (Section 5.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation (Section 5.0). The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow procedures which resulted in an inadequate operability evaluation. Specifically, the evaluation did not include an analysis of conditions that could be causing the valve to fail, and it provided no assessment of the effect these conditions would have related to the specified safety function and mission time of the standby diesel generator. The licensee entered this issue in their corrective action program as Condition Report GGN-2007-2256. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions (Section 5.0). Inspection Report# : 2007006 (pdf) Significance: Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Station Procedures for Conducting Maintenance Activities The inspectors reviewed a Green, self-revealing noncited violation of Technical Specification 5.4.1(a) for failure to follow station maintenance procedures while troubleshooting the control rod drive Pump A hand switch green indicating light socket. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-4474. The finding is more than minor since it affects the human performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, improper maintenance practices on control room equipment could lead to a more significant safety concern. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, inspectors determined that the finding has very low safety significance because it did not result in a loss of safety function. This finding has a crosscutting aspect in the area of human performance associated with work practices in that licensee personnel proceeded to troubleshoot the bulb in the face of uncertainty surrounding the required bulb type and expected system response (Section 1R19).

Inspection Report# : 2006005 (pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Loose Items in Safety Related Areas The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for a failure to control loose items in safety related structures. Between July 25 and September 13, 2006, the inspectors identified six examples of loose items in the auxiliary building and control building that did not meet the requirements of plant loose item control procedures. The licensee entered this issue in their corrective action program as CR-GGN-2006-3836. The failure to control loose items in the vicinity of safety related equipment was a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of protection against external factors (seismic) and affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability. The cause of this finding is related to the cross-cutting element of human performance in that licensee work practices did not effectively define and communicate expectations regarding compliance with plant procedures for the control of loose items in safety related structures. Inspection Report# : 2006004 (pdf) Significance: Aug 07, 2006 Identified By: Self-Revealing Item Type: NCV NonCited Violation Standby Diesel Generator Cylinder Head Failures A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for the failure of licensee personnel to preclude repetition of a significant condition adverse to quality. Specifically, the licensee failed to take actions to prevent subsequent standby diesel generator engine head failures attributed to corrosion fatigue in 1992, 1996, and 2006. This issue was entered into the licensee's corrective action program as Conditon Report CR-GGN-2006-1955. The finding was more than minor since it affected the Mitigation System Cornerstone attribute of availability and reliability of mitigating equipment, specifically the standby diesel generators. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding is of very low significance since it only involved the loss of one train of diesel generators for less than the technical specification allowed outage time (Section 4.b). Inspection Report# : 2006010 (pdf) Barrier Integrity Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control of Leakage Detection Sensing Lines The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to maintain the original design configuration of a leakage detection instrument sensing line in the fuel pool cooling and cleanup system. The licensee entered this issue in their corrective action program as CR-GGN-2006-3569. This finding is more than minor since it affects the design control attribute of the spent fuel pool cooling aspect of the Barrier Integrity cornerstone and affects the cornerstone objective of providing assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it only affected the radiological barrier function provided by the spent fuel pool.

Inspection Report# : 2006004 (pdf) Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Containment Pool Liner Leakage The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure of the licensee to take actions required by operator rounds in response to containment pool liner leakage. The licensee entered this issue into their corrective action program as CR-GGN-2006-3500. The finding was more than minor since the failure of operators to perform operator rounds could lead to a more significant safety concern if left uncorrected. Additionally, the identified liner leakage represented a degrading condition that, if left uncorrected, could continue to degrade and could potentially result in the migration of water to other portions of the containment structure. The inspectors determined this finding affected the Barrier Integrity cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it does not represent an actual open pathway in the physical integrity of the reactor containment or an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. The cause of this finding is related to the crosscutting element of human performance in that licensee work practices did not effectively define and communicate expectations regarding compliance with plant procedures for the conduct of operator rounds. Inspection Report# : 2006004 (pdf) Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedural Guidance and Radiation Work Instructions While Supporting Radiograpy Operations A self-revealing, Green noncited violation of Technical Specification 5.4.1 was identified for the failure to follow procedural guidance and radiation work instructions while supporting radiography operations. All entrances to the area in which radiography was conducted were not barricaded and posted at the two millirem per hour point, as required. However, the high radiation area was barricaded, posted, and guarded. As immediate corrective action, the licensee postponed additional radiography and initiated a review of the occurrence. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective, in that the failure to control access to areas in which radiography is conducted could result in unplanned personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety signficance because (1) it was not an ALARA finding, (2) there was no overexposeure, (3) there was no substantial potential for an overexposure because no one entered the area in which high doses were possible, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work control because the licensee did not coordinate work activities by incorporating actions to address the need to keep personnel apprised of work status. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation

Failure to Follow Radiation Work Permit Instructions Prohibiting High Radiation Area Entry. A self-revealing Green noncited violation of Technical Specification 5.4.1 was identified for the failure to follow radiation work permit instructions prohibiting high radiation area entry. Two outage workers entered a high radiation area on the 139-foot elevation of the auxiliary steam tunnel, in violation of their radiation work permit instructions. The licensee was alerted to the entry into the high radiation area by one of the workers alarming dosimeter. As immediate corrective action, the licensee revoked the worker's access to the radiologically controlled area. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective, in that the failure to follow radiation work permit instructions could result in unplanned personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety significance because (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for overexposure because, at the highest dose rate, it would have taken 40 hours to receive a whole-body overexposure, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work practices because the workers failed to use error prevention techniques such as self and peer checking. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Radiological Hazard Caused by Water Leaking in the Drywell. The inspector identified a Green noncited violation of 10 CFR 20.1501(a) because the licensee failed to adequately evaluate the radiological hazard caused by water leaking from a valve in the drywell. The licensee failed to maintain knowledge of changing radiological conditions. As immediate corrective action, the licensee surveyed the area to obtain current information. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety significance because (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions in deciding the correct contamination survey frequency in the drywell. Inspection Report# : 2007002 (pdf) Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : August 24, 2007

Grand Gulf 1 3Q/2007 Plant Inspection Findings Initiating Events Significance: Dec 31, 2006 Identified By: NRC Item Type: FIN Finding Insufficient Preventive Maintenance of Bus Duct Cooling System Results in Unplanned Power Reduction The inspectors reviewed a Green, self-revealing finding for failure to implement preventive maintenance on the bus duct cooling system components prior to system failures, causing a plant transient. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-3996. The finding is more than minor since it affects the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding has a very low safety significance since it did not contribute to the likelihood of a loss of coolant accident, did not contribute to a loss of mitigation equipment, and did not increase the likelihood of a fire or internal/external flood (Section 4OA3). Inspection Report# : 2006005 (pdf) Mitigating Systems Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Leakage Detection System The inspectors identified a Green noncited violation involving the failure to adequately monitor the performance of the leakage detection system in accordance with 10CFR50.65(a)(2). Specifically, the licensee failed to account for the functional failure of a temperature switch which resulted in exceeding the performance criteria for the leakage detection system. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-2955. This finding was greater than minor since violations of 10 CFR Part 50.65(a)(2) necessarily involve degraded system performance which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the maintenance rule aspect of the finding did not lead to an actual loss of safety function of the system nor did it cause a component to be inoperable. This finding has a crosscutting aspect in the area of human performance associated with work practices in that the licensee failed to use human error prevention techniques such as self checking and peer checking when utilizing the maintenance rule database (H.4(a)). Inspection Report# : 2007003 (pdf) Significance: Jun 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Foreign Material Controls During Reactor Feed Pump Maintenance A self-revealing Green finding was identified for inadequate foreign material controls during maintenance. Specifically, a foreign material exclusion device was left inside the reactor feed Pump B lube oil system following maintenance activities, which prevented placing the pump in service during reactor startup. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-2158. The finding was more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of

systems that respond to initiating events. The inspectors determined this finding required a Phase 2 analysis because it resulted in the loss of function of a single train of the power conversion system (reactor feed) for greater than 24 hours. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance because of the availability of the condensate booster pumps and emergency core cooling systems. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources because licensee personnel were not adequately trained to consistently implement the foreign material exclusion program (H.2(b)). Inspection Report# : 2007003 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure for Safety-Related Breaker Inspections The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure to identify loose and missing fasteners on the standby service water Train B bus feeder breaker. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-3081. This finding was more than minor because the failure to ensure that loose parts are not present in safety related breakers, if left uncorrected, could become a more significant safety concern. Using the Significance Determination Process Phase 1 Screening Worksheet in Appendix A of Inspection Manual Chapter 0609, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. Inspection Report# : 2007003 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation. The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving the failure to follow procedures resulted in an inadequate operability evaluation for a degraded switchgear ventilation system. Specifically, the evaluation utilized several non-conservative input assumptions and failed to adequately evaluate the potential adverse affects from changing weather conditions. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-0554. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it did not result in a loss of operability. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions (Section 1R15). Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Command and Control Results in Inappropriate Valve Manipulations. The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for failure to meet procedural requirements involving command and control in the control room. Specifically, the control room supervisor was not informed of a system alignment change directed by the shift technical advisor. The licensee entered this issue in their corrective action program as CR-GGN-2007-1060. This finding is more than minor since the failure to maintain appropriate command and control in the control room, if left uncorrected, could lead to a more significant safety concern. The inspectors determined that this finding affected the mitigating systems cornerstone. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, the finding is of very low safety significance since it did not result in an actual loss of operability. This finding has a crosscutting aspect in the area of human performance associated with work practices because the failure to communicate the system realignment to the control room supervisor prevented the control room supervisor from maintaining proper supervisory oversight of work activities. Inspection Report# : 2007002 (pdf)

Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Resulting in Isolation of Switchgear Room Ventilation. A self-revealing Green noncited violation of Technical Specification 5.4.1(a) was identified for the failure to follow a surveillance procedure resulting in the inadvertent isolation of ventilation to the Division 1 and Division 3 safety-related switchgear rooms. The licensee entered this issue in their corrective action program as CR-GGN-2006-4394. This finding is more than minor since it affected the human performance attribute of the mitigating systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. This finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel did not effectively utilize human error prevention techniques, such as self and peer checking. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Standby Service Water System Leakage. The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee failed to take adequate corrective actions in response to service water leakage from drywell purge compressor oil cooler drain plugs. The licensee entered this issue in their corrective action program as CR-GGN-2006-4762. This finding is more than minor because if left uncorrected, the zinc drain plugs could have deteriorated to a point at which service water leakage would have impacted the performance of the standby service water system. This finding also affects the equipment performance attribute of the mitigating systems cornerstone and impacts the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Using the Significance Determination Process Phase 1 Screening Worksheet in Appendix A of Inspection Manual Chapter 0609, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to thoroughly evaluate the cause and extent of condition for corrosion identified on the drain plugs of the Train B purge compressor oil cooler. Inspection Report# : 2007002 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of High Standby Diesel Generator Temperatures (Section 3.0). The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to identify and correct the cause of elevated temperatures adversely affecting the safety function of the Division I standby diesel generator that had previously occurred in 1999 and 2004. Subsequently, on January 30, 2007, the Division I standby diesel generator again experienced elevated temperatures during a surveillance run and was subsequently declared inoperable. This issue was entered into the licensee's corrective action program as Condition Report GGN-2007-0378. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The Phase 1 Worksheets in Manual Chapter 0609, Significance Determination Process, were used to conclude that a Phase 2 analysis was required because the condition represented a loss of safety function of a single train of a Technical Specification system for greater than its allowed outage time. The inspectors performed a Phase 2 analysis using Appendix A, Technical Basis For At Power Significance Determination Process, of Manual Chapter 0609, Significance Determination Process, and the Phase 2 Worksheet for Grand Gulf. The Phase 2 evaluation concluded that the finding was of very low safety significance. A Phase 3 significance determination analysis also determined the finding was of very low safety significance. The cause of the finding is related to the problem identification and resolution crosscutting area in that

the licensee failed to thoroughly evaluate the problem resulting in ineffective corrective actions being implemented that failed to prevent recurrence of a significant condition adverse to quality (Section 3.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Alarm Response Instruction for SDG High Jacket Water Temperature (Section 4.0). The team identified a noncited violation of Technical Specification 5.4.1 (a) involving the failure to maintain an adequate alarm response instruction for standby diesel generator high jacket water temperature. Specifically, Procedure 04-1-02-1H22-P400, Alarm Response Instruction, Panel No.: 1H-22-P400, Safety Related, Revision 109, failed to provide adequate guidance to manually override the standby diesel generator jacket water cooling system temperature control valve during emergency conditions. This issue was entered into the licensee's corrective action program as Condition Report GG-2007-1837. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of procedure quality and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance because it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating events. The cause of the finding is related to the problem identification and resolution crosscutting area in that the licensee did not take appropriate corrective actions to adequately address a previously identified safety concern (Section 4.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify a Degraded Condition (Section 5.0). The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to promptly identify a condition adverse to quality. Between February 2-15, 2007, the licensee failed to promptly identify that corrective actions taken in response to a January 30, 2007, failure of the Division 1 standby diesel generator jacket water cooling system temperature control valve had not addressed the cause of the valve failure. Specifically, following the valves failure, the licensee inappropriately concluded the valves internal thermal elements were faulty, replaced the elements, performed postmaintenance testing, and declared the valve and associated standby diesel generator operable on February 1, 2007. Subsequent testing of the suspect faulty thermal elements on February 2 and 13, 2007, found the components were functional. Following receipt of the testing results, the licensee failed to promptly identify that replacement of the thermal elements failed to address the cause of the problem resulting in the failure to evaluate a potential degraded condition affecting operability of the standby emergency diesel generator. This issue was entered into the licensee's corrective action program as Condition Report GGN-2007-2255. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associate cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance because the condition did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating events. The cause of the finding is related to the problem identification and resolution crosscutting area in that the licensee did not identify an issue completely, accurately, and in a timely manner commensurate with its safety significance resulting in the failure to evaluate a potential degraded condition for operability (Section 5.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation (Section 5.0). The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow procedures which resulted in an inadequate operability evaluation.

Specifically, the evaluation did not include an analysis of conditions that could be causing the valve to fail, and it provided no assessment of the effect these conditions would have related to the specified safety function and mission time of the standby diesel generator. The licensee entered this issue in their corrective action program as Condition Report GGN-2007-2256. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions (Section 5.0). Inspection Report# : 2007006 (pdf) Significance: Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Station Procedures for Conducting Maintenance Activities The inspectors reviewed a Green, self-revealing noncited violation of Technical Specification 5.4.1(a) for failure to follow station maintenance procedures while troubleshooting the control rod drive Pump A hand switch green indicating light socket. The licensee entered this into their corrective action program as Condition Report CR-GGN-2006-4474. The finding is more than minor since it affects the human performance attribute of the mitigating systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, improper maintenance practices on control room equipment could lead to a more significant safety concern. Using the NRC Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, inspectors determined that the finding has very low safety significance because it did not result in a loss of safety function. This finding has a crosscutting aspect in the area of human performance associated with work practices in that licensee personnel proceeded to troubleshoot the bulb in the face of uncertainty surrounding the required bulb type and expected system response (Section 1R19). Inspection Report# : 2006005 (pdf) Barrier Integrity Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Repair Crack in Containment Building Structure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving the failure to correct a crack in the ceiling of the reactor water cleanup heat exchanger room internal to the containment building structure. Specifically, the licensee identified the crack in 1987 but failed to complete planned corrective actions to evaluate or repair the crack during Refueling Outage 2. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-1970. This finding was more than minor because the reactor water cleanup (RWCU) ceiling crack represented a degrading condition that if left uncorrected could become a more significant safety concern. The inspectors determined this finding affected the Barrier Integrity cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not represent an actual open pathway in the physical integrity of the reactor containment or an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. Inspection Report# : 2007003 (pdf) Emergency Preparedness

Occupational Radiation Safety Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Radiological Hazard Caused by Foreign Material Retrieval from the Reactor Vessel The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 20.1501(a) because the licensee failed to evaluate the radiological hazard of foreign material retrieval from the reactor vessel. A contract radiation protection technician misinterpreted his survey instrument readings, picked up a bolt with a radiation dose rate of 19.9 rem per hour, and received a shallow dose equivalent of 41 millirems. The radiation protection technician was alerted to the problem by an electronic dosimeter alarm. As corrective action, the licensee revised the appropriate radiation work permit template to incorporate a dose rate limit for items removed from pools and included a discussion of the violation in radiation protection training. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable (ALARA) planning or work control issue; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work practices because the workers failed to use error prevention techniques such as peer checking and self checking (H.4(a)). Inspection Report# : 2007003 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post and Control a High Radiation Area The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.7.1 resulting from a failure to post and control a high radiation area. Room 0R123 on the 93-foot elevation of the radwaste building had dose rates as high as 265 millirems per hour at 30 centimeters from the G17D069 filter housing and was not posted and controlled as a high radiation area. The licensee was alerted to the situation when the electronic dosimeters of two radwaste operators alarmed when they entered the higher dose rates. Poor communications between operations and radiation protection personnel contributed to the failure to identify the high radiation area. Radiation protection supervisors stated they were unaware at the time of the operators dose rate alarms that reactor water cleanup reject flow was approximately twice the normal flow rate and both of the reactor water cleanup demineralizers had been out of service from approximately 3:00 p.m. on May 19 until 9:00 a.m. on May 20, 2007. As immediate corrective action, the area was barricaded and conspicuously posted as a high radiation area. Additional planned corrective actions were still being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the failure to post and control a high radiation area had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable (ALARA) planning or work control issue; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work control because the licensee failed to ensure proper communication, coordination, and cooperation during activities in which interdepartmental coordination was necessary to assure plant and human performance (H.3(b)). Inspection Report# : 2007003 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation

Failure to Follow Procedural Guidance and Radiation Work Instructions While Supporting Radiograpy Operations A self-revealing, Green noncited violation of Technical Specification 5.4.1 was identified for the failure to follow procedural guidance and radiation work instructions while supporting radiography operations. All entrances to the area in which radiography was conducted were not barricaded and posted at the two millirem per hour point, as required. However, the high radiation area was barricaded, posted, and guarded. As immediate corrective action, the licensee postponed additional radiography and initiated a review of the occurrence. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective, in that the failure to control access to areas in which radiography is conducted could result in unplanned personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety signficance because (1) it was not an ALARA finding, (2) there was no overexposeure, (3) there was no substantial potential for an overexposure because no one entered the area in which high doses were possible, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work control because the licensee did not coordinate work activities by incorporating actions to address the need to keep personnel apprised of work status. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit Instructions Prohibiting High Radiation Area Entry. A self-revealing Green noncited violation of Technical Specification 5.4.1 was identified for the failure to follow radiation work permit instructions prohibiting high radiation area entry. Two outage workers entered a high radiation area on the 139-foot elevation of the auxiliary steam tunnel, in violation of their radiation work permit instructions. The licensee was alerted to the entry into the high radiation area by one of the workers alarming dosimeter. As immediate corrective action, the licensee revoked the worker's access to the radiologically controlled area. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective, in that the failure to follow radiation work permit instructions could result in unplanned personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety significance because (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for overexposure because, at the highest dose rate, it would have taken 40 hours to receive a whole-body overexposure, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work practices because the workers failed to use error prevention techniques such as self and peer checking. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Radiological Hazard Caused by Water Leaking in the Drywell. The inspector identified a Green noncited violation of 10 CFR 20.1501(a) because the licensee failed to adequately evaluate the radiological hazard caused by water leaking from a valve in the drywell. The licensee failed to maintain knowledge of changing radiological conditions. As immediate corrective action, the licensee surveyed the area to obtain current information. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety significance because (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions in deciding the correct contamination survey frequency in the drywell.

Inspection Report# : 2007002 (pdf) Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : December 07, 2007

Grand Gulf 1 4Q/2007 Plant Inspection Findings Initiating Events Significance: Sep 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Reactor SCRAM due to Turbine Trip caused by Loss of Condenser Vacuum A self-revealing finding was identified involving the failure to properly calibrate the main condenser hydraulic vacuum switch that established a higher trip setpoint that would prematurely actuate an automatic turbine trip and reactor scram for a degraded main condenser vacuum condition. This issue was entered into the licensees corrective action program as condition Report CR-GGN-2007-02756. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the MC 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance, because the finding did not contribute to the likelihood that mitigating equipment or functions would not be available following a reactor trip. The cause of the finding was related to the human performance crosscutting component of resources in that the calibration procedure did not provide clear instructions detailing the methodology to adjust the speed simulation screw to the required position.[H.2(c)] Inspection Report# : 2007004 (pdf) Mitigating Systems Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of the Control Rod Drive System. The inspectors identified a Green noncited violation (NCV) of 10 CFR Part 50.65(a)(2) for the failure to adequately monitor the performance of the control rod drive system. Specifically, the licensee failed to adequately perform a functional failure determination for a degraded flow control valve. Following licensee review of this condition the system was placed in the maintenance rule (a)(1) monitoring status. This finding was more than minor since the degraded control rod drive flow control valve caused the system to be placed in the (a)(1) monitoring status. This finding was characterized under the significance determination process as having a very low safety significance, because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system, nor did it cause a component to become inoperable. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making, because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions. H.1(b) Inspection Report# : 2007004 (pdf) Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish a Formal Procedure to Monitor Outdoor Air Temperatures. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the

failure to translate a design basis limit for outdoor air temperature into an instruction or procedure. The licensee established a new Updated Final Safety Analysis Report maximum outdoor air temperature of 102.5 degrees F. If outside air temperatures exceeded 102.5 degrees F, safety-related equipment, which are located in rooms that are cooled by outdoor air (i.e., standby service water pump room), would be operationally challenged. The inspectors identified that no instruction or procedure was established to monitor high outside temperature or subsequent actions established in the event the design basis temperature limit is exceeded. The inspectors determined that the finding was more than minor because the finding affects the mitigating system cornerstone objective of ensuring the reliability of the standby service water system that responds to initiating events to prevent undesirable conditions. Using the Phase 1 worksheet in Inspection Manual Chapter 0609, "Significance Determination Process," this finding is determined to be of very low safety significance because there was no actual loss of a safety function, and the design basis limits had not been exceeded. The inspectors determined that the finding has a crosscutting aspect in the area of human performance decision making because the licensee failed to use conservative assumptions in determining not to establish a procedure or instruction to monitor high outside temperature for design limits on the standby service water pump room. H.1(b) Inspection Report# : 2007004 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Leakage Detection System The inspectors identified a Green noncited violation involving the failure to adequately monitor the performance of the leakage detection system in accordance with 10CFR50.65(a)(2). Specifically, the licensee failed to account for the functional failure of a temperature switch which resulted in exceeding the performance criteria for the leakage detection system. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-2955. This finding was greater than minor since violations of 10 CFR Part 50.65(a)(2) necessarily involve degraded system performance which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the maintenance rule aspect of the finding did not lead to an actual loss of safety function of the system nor did it cause a component to be inoperable. This finding has a crosscutting aspect in the area of human performance associated with work practices in that the licensee failed to use human error prevention techniques such as self checking and peer checking when utilizing the maintenance rule database (H.4(a)). Inspection Report# : 2007003 (pdf) Significance: Jun 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Foreign Material Controls During Reactor Feed Pump Maintenance A self-revealing Green finding was identified for inadequate foreign material controls during maintenance. Specifically, a foreign material exclusion device was left inside the reactor feed Pump B lube oil system following maintenance activities, which prevented placing the pump in service during reactor startup. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-2158. The finding was more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The inspectors determined this finding required a Phase 2 analysis because it resulted in the loss of function of a single train of the power conversion system (reactor feed) for greater than 24 hours. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance because of the availability of the condensate booster pumps and emergency core cooling systems. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources because licensee personnel were not adequately trained to consistently implement the foreign material exclusion program (H.2(b)). Inspection Report# : 2007003 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation

Failure to Follow Procedure for Safety-Related Breaker Inspections The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure to identify loose and missing fasteners on the standby service water Train B bus feeder breaker. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-3081. This finding was more than minor because the failure to ensure that loose parts are not present in safety related breakers, if left uncorrected, could become a more significant safety concern. Using the Significance Determination Process Phase 1 Screening Worksheet in Appendix A of Inspection Manual Chapter 0609, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. Inspection Report# : 2007003 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation. The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving the failure to follow procedures resulted in an inadequate operability evaluation for a degraded switchgear ventilation system. Specifically, the evaluation utilized several non-conservative input assumptions and failed to adequately evaluate the potential adverse affects from changing weather conditions. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-0554. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding is of very low safety significance since it did not result in a loss of operability. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions (Section 1R15). Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Command and Control Results in Inappropriate Valve Manipulations. The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for failure to meet procedural requirements involving command and control in the control room. Specifically, the control room supervisor was not informed of a system alignment change directed by the shift technical advisor. The licensee entered this issue in their corrective action program as CR-GGN-2007-1060. This finding is more than minor since the failure to maintain appropriate command and control in the control room, if left uncorrected, could lead to a more significant safety concern. The inspectors determined that this finding affected the mitigating systems cornerstone. Using Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, the finding is of very low safety significance since it did not result in an actual loss of operability. This finding has a crosscutting aspect in the area of human performance associated with work practices because the failure to communicate the system realignment to the control room supervisor prevented the control room supervisor from maintaining proper supervisory oversight of work activities. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Resulting in Isolation of Switchgear Room Ventilation. A self-revealing Green noncited violation of Technical Specification 5.4.1(a) was identified for the failure to follow a surveillance procedure resulting in the inadvertent isolation of ventilation to the Division 1 and Division 3 safety-related switchgear rooms. The licensee entered this issue in their corrective action program as CR-GGN-2006-4394. This finding is more than minor since it affected the human performance attribute of the mitigating systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined the finding was of very low safety significance because it did not result in a loss

of operability. This finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel did not effectively utilize human error prevention techniques, such as self and peer checking. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Standby Service Water System Leakage. The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee failed to take adequate corrective actions in response to service water leakage from drywell purge compressor oil cooler drain plugs. The licensee entered this issue in their corrective action program as CR-GGN-2006-4762. This finding is more than minor because if left uncorrected, the zinc drain plugs could have deteriorated to a point at which service water leakage would have impacted the performance of the standby service water system. This finding also affects the equipment performance attribute of the mitigating systems cornerstone and impacts the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Using the Significance Determination Process Phase 1 Screening Worksheet in Appendix A of Inspection Manual Chapter 0609, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to thoroughly evaluate the cause and extent of condition for corrosion identified on the drain plugs of the Train B purge compressor oil cooler. Inspection Report# : 2007002 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of High Standby Diesel Generator Temperatures (Section 3.0). The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to identify and correct the cause of elevated temperatures adversely affecting the safety function of the Division I standby diesel generator that had previously occurred in 1999 and 2004. Subsequently, on January 30, 2007, the Division I standby diesel generator again experienced elevated temperatures during a surveillance run and was subsequently declared inoperable. This issue was entered into the licensee's corrective action program as Condition Report GGN-2007-0378. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The Phase 1 Worksheets in Manual Chapter 0609, Significance Determination Process, were used to conclude that a Phase 2 analysis was required because the condition represented a loss of safety function of a single train of a Technical Specification system for greater than its allowed outage time. The inspectors performed a Phase 2 analysis using Appendix A, Technical Basis For At Power Significance Determination Process, of Manual Chapter 0609, Significance Determination Process, and the Phase 2 Worksheet for Grand Gulf. The Phase 2 evaluation concluded that the finding was of very low safety significance. A Phase 3 significance determination analysis also determined the finding was of very low safety significance. The cause of the finding is related to the problem identification and resolution crosscutting area in that the licensee failed to thoroughly evaluate the problem resulting in ineffective corrective actions being implemented that failed to prevent recurrence of a significant condition adverse to quality (Section 3.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Alarm Response Instruction for SDG High Jacket Water Temperature (Section 4.0). The team identified a noncited violation of Technical Specification 5.4.1 (a) involving the failure to maintain an adequate alarm response instruction for standby diesel generator high jacket water temperature. Specifically, Procedure 04-1-02-1H22-P400, Alarm Response Instruction, Panel No.: 1H-22-P400, Safety Related, Revision 109,

failed to provide adequate guidance to manually override the standby diesel generator jacket water cooling system temperature control valve during emergency conditions. This issue was entered into the licensee's corrective action program as Condition Report GG-2007-1837. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of procedure quality and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance because it did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating events. The cause of the finding is related to the problem identification and resolution crosscutting area in that the licensee did not take appropriate corrective actions to adequately address a previously identified safety concern (Section 4.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Identify a Degraded Condition (Section 5.0). The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to promptly identify a condition adverse to quality. Between February 2-15, 2007, the licensee failed to promptly identify that corrective actions taken in response to a January 30, 2007, failure of the Division 1 standby diesel generator jacket water cooling system temperature control valve had not addressed the cause of the valve failure. Specifically, following the valves failure, the licensee inappropriately concluded the valves internal thermal elements were faulty, replaced the elements, performed postmaintenance testing, and declared the valve and associated standby diesel generator operable on February 1, 2007. Subsequent testing of the suspect faulty thermal elements on February 2 and 13, 2007, found the components were functional. Following receipt of the testing results, the licensee failed to promptly identify that replacement of the thermal elements failed to address the cause of the problem resulting in the failure to evaluate a potential degraded condition affecting operability of the standby emergency diesel generator. This issue was entered into the licensee's corrective action program as Condition Report GGN-2007-2255. The finding is greater than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and affects the associate cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance because the condition did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating events. The cause of the finding is related to the problem identification and resolution crosscutting area in that the licensee did not identify an issue completely, accurately, and in a timely manner commensurate with its safety significance resulting in the failure to evaluate a potential degraded condition for operability (Section 5.0). Inspection Report# : 2007006 (pdf) Significance: Mar 12, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation (Section 5.0). The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for a failure to follow procedures which resulted in an inadequate operability evaluation. Specifically, the evaluation did not include an analysis of conditions that could be causing the valve to fail, and it provided no assessment of the effect these conditions would have related to the specified safety function and mission time of the standby diesel generator. The licensee entered this issue in their corrective action program as Condition Report GGN-2007-2256. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions (Section 5.0). Inspection Report# : 2007006 (pdf)

Barrier Integrity Significance: Sep 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures Caused Loss of Decay Heat Removal in the Spent Fuel Pool. A self-revealing Green non-cited violation of Technical Specifications 5.4.1(a) was identified involving the failure to adequately follow procedure to align valves in the fuel pool cooling and cleanup system. The valves were aligned in the wrong sequence, contrary to the system operating instructions, causing both fuel pool cooling and cleanup pumps to trip and a subsequent loss of fuel pool cooling. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-04284. The finding is more than minor, since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has a very low safety significance since it only represents a degradation of the radiological barrier function provided by the spent fuel pool system. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel failed to follow the correct sequence of valve manipulations required by procedure. [H.4 (b)] Inspection Report# : 2007004 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Repair Crack in Containment Building Structure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving the failure to correct a crack in the ceiling of the reactor water cleanup heat exchanger room internal to the containment building structure. Specifically, the licensee identified the crack in 1987 but failed to complete planned corrective actions to evaluate or repair the crack during Refueling Outage 2. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-1970. This finding was more than minor because the reactor water cleanup (RWCU) ceiling crack represented a degrading condition that if left uncorrected could become a more significant safety concern. The inspectors determined this finding affected the Barrier Integrity cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not represent an actual open pathway in the physical integrity of the reactor containment or an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. Inspection Report# : 2007003 (pdf) Emergency Preparedness Occupational Radiation Safety Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Radiological Hazard Caused by Foreign Material Retrieval from the Reactor Vessel The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 20.1501(a) because the licensee failed to

evaluate the radiological hazard of foreign material retrieval from the reactor vessel. A contract radiation protection technician misinterpreted his survey instrument readings, picked up a bolt with a radiation dose rate of 19.9 rem per hour, and received a shallow dose equivalent of 41 millirems. The radiation protection technician was alerted to the problem by an electronic dosimeter alarm. As corrective action, the licensee revised the appropriate radiation work permit template to incorporate a dose rate limit for items removed from pools and included a discussion of the violation in radiation protection training. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable (ALARA) planning or work control issue; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work practices because the workers failed to use error prevention techniques such as peer checking and self checking (H.4(a)). Inspection Report# : 2007003 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post and Control a High Radiation Area The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.7.1 resulting from a failure to post and control a high radiation area. Room 0R123 on the 93-foot elevation of the radwaste building had dose rates as high as 265 millirems per hour at 30 centimeters from the G17D069 filter housing and was not posted and controlled as a high radiation area. The licensee was alerted to the situation when the electronic dosimeters of two radwaste operators alarmed when they entered the higher dose rates. Poor communications between operations and radiation protection personnel contributed to the failure to identify the high radiation area. Radiation protection supervisors stated they were unaware at the time of the operators dose rate alarms that reactor water cleanup reject flow was approximately twice the normal flow rate and both of the reactor water cleanup demineralizers had been out of service from approximately 3:00 p.m. on May 19 until 9:00 a.m. on May 20, 2007. As immediate corrective action, the area was barricaded and conspicuously posted as a high radiation area. Additional planned corrective actions were still being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the failure to post and control a high radiation area had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable (ALARA) planning or work control issue; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work control because the licensee failed to ensure proper communication, coordination, and cooperation during activities in which interdepartmental coordination was necessary to assure plant and human performance (H.3(b)). Inspection Report# : 2007003 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedural Guidance and Radiation Work Instructions While Supporting Radiograpy Operations A self-revealing, Green noncited violation of Technical Specification 5.4.1 was identified for the failure to follow procedural guidance and radiation work instructions while supporting radiography operations. All entrances to the area in which radiography was conducted were not barricaded and posted at the two millirem per hour point, as required. However, the high radiation area was barricaded, posted, and guarded. As immediate corrective action, the licensee postponed additional radiography and initiated a review of the occurrence. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective, in that the failure to control access to areas in which radiography is conducted could result in unplanned personnel dose. Using the Occupational Radiation Safety

Significance Determination Process, the inspector determined the finding had very low safety signficance because (1) it was not an ALARA finding, (2) there was no overexposeure, (3) there was no substantial potential for an overexposure because no one entered the area in which high doses were possible, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work control because the licensee did not coordinate work activities by incorporating actions to address the need to keep personnel apprised of work status. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Radiation Work Permit Instructions Prohibiting High Radiation Area Entry. A self-revealing Green noncited violation of Technical Specification 5.4.1 was identified for the failure to follow radiation work permit instructions prohibiting high radiation area entry. Two outage workers entered a high radiation area on the 139-foot elevation of the auxiliary steam tunnel, in violation of their radiation work permit instructions. The licensee was alerted to the entry into the high radiation area by one of the workers alarming dosimeter. As immediate corrective action, the licensee revoked the worker's access to the radiologically controlled area. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program attribute of exposure control and affected the cornerstone objective, in that the failure to follow radiation work permit instructions could result in unplanned personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety significance because (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for overexposure because, at the highest dose rate, it would have taken 40 hours to receive a whole-body overexposure, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work practices because the workers failed to use error prevention techniques such as self and peer checking. Inspection Report# : 2007002 (pdf) Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Radiological Hazard Caused by Water Leaking in the Drywell. The inspector identified a Green noncited violation of 10 CFR 20.1501(a) because the licensee failed to adequately evaluate the radiological hazard caused by water leaking from a valve in the drywell. The licensee failed to maintain knowledge of changing radiological conditions. As immediate corrective action, the licensee surveyed the area to obtain current information. Further corrective action is being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined the finding had very low safety significance because (1) it was not an ALARA finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. Additionally, this finding has a crosscutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions in deciding the correct contamination survey frequency in the drywell. Inspection Report# : 2007002 (pdf) Public Radiation Safety Significance: Sep 14, 2007 Identified By: NRC Item Type: NCV NonCited Violation

Failure to Provide Function-Specific Training to Hazardous Material Workers. The team identified a noncited violation (NCV) of 10 CFR 71.5 because the licensee failed to provide required training to hazardous material workers involved in the shipment of radioactive material. Specifically, the licensee did not provide function-specific training, pursuant to 49 CFR 172.704(a) of Department of Transportation shipping regulations, to maintenance personnel involved in the reassembly the shipping casks. Corrective actions are still being evaluated; however, the licensee plans to provide hazardous material training to these employees. The licensee documented this issue in the corrective action program as Condition Report-GGN-2007-04572. The finding is greater than minor because it is associated with the Public Radiation Safety Cornerstone attribute of program and process and affects the cornerstone objective. Inadequate training of hazardous material workers regarding the reassembly and loading of shipping casks has a potential impact on public dose and on the licensees ability to safely package and transport radioactive material on public roadways. The violation involved an occurrence in the licensees radioactive material transportation program that is contrary to NRC or Department of Transportation regulations. When processed through the Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it: (1) was associated with radioactive material control, (2) involved the licensees program for radioactive material packaging and transportation, (3) did not cause radiation limits to be exceeded, (4) did not result in a breach of package during transit, (5) did not involve a certificate of compliance issue, (6) did not involve a non-compliance with low level burial ground, and (7) did not involve a failure to make notifications or to provide emergency information. In addition, this finding had cross-cutting aspects in the area of human performance in the component of resources because the licensee did not ensure the availability and adequacy of training for hazardous material workers involved in the shipment of radioactive material was available. Hazardous material workers are required to be trained and qualified to prepare radioactive material shipments for transport. (H.2.b) (Section 2PS2) Inspection Report# : 2007007 (pdf) Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : February 04, 2008

Grand Gulf 1 1Q/2008 Plant Inspection Findings Initiating Events Significance: Mar 22, 2008 Identified By: NRC Item Type: FIN Finding Ineffective Corrective Actions in Response to Resin in the Electro-hydraulic Control System. The inspectors identified a finding involving ineffective corrective actions in response to resin intrusion in the electro-hydraulic control system. The inspectors reviewed the corrective actions from a condition report involving a resin intrusion into the electro-hydraulic control system via a failed temporary ion-exchange filter in 2003. Review of the corrective actions associated with the 2003 event revealed that a long-range recovery plan was developed to remove resin from the electro-hydraulic control system. However, the recovery plan corrective actions were closed without licensee actions to remove resin from the electro-hydraulic control system. The failure to implement effective corrective actions following the 2003 resin intrusion event directly resulted in electro-hydraulic control stability issues seen in the fall of 2007, including reactor pressure perturbations and reductions in reactor power. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2007-04972. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 worksheet, the finding was determined to have very low safety significance because the finding did not contribute to the likelihood that mitigating equipment would not be available following a reactor trip. (Section 4OA2) Inspection Report# : 2008002 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Improper Control of Troubleshooting Causes a Loss of Condenser Vacuum The inspectors identified a finding involving a loss of condenser vacuum caused by improper troubleshooting of the seal steam pressure controller. Specifically, the licensee failed to provide adequate work instructions and procedural limitations during troubleshooting of the seal steam pressure controller. As a result, the plant experienced a loss of condenser vacuum and a plant transient. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-04626. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the initiating event and mitigating systems cornerstone. The inspectors performed a Phase 2 analysis using Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," of Manual Chapter 0609, "Significance Determination Process," and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors assumed that only the power conversion system was affected and all other mitigating systems were available. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding was related to the human performance crosscutting component of decision making, in that the licensee failed to use conservative assumptions during troubleshooting activities and performed these activities without determining the validity of the troubleshooting instructions and identifying possible unintended consequences [H.1(b)] (Section 4OA3). Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Engineering Review of Plant Service Water Modification A self-revealing finding was identified involving the failure of a plant service water piping flange due to an improper flow control valve design modification. Specifically, the licensee failed to perform an adequate review of an engineering modification and the maintenance work orders did not have detailed installation instructions. As a result, the plant experienced a plant transient. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-05040. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because the finding did not contribute to the likelihood that mitigating equipment or functions would not be available following a reactor trip. The cause of the finding was related to the human performance crosscutting component of work practices in that the responsible engineers failed to perform adequate self and peer checking during the development and review of the design modification to the plant service water flow control check valves [H.4(a)] (Section 4OA3). Inspection Report# : 2007005 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Follow Procedure Results in Loss of Condener Vacuum A self-revealing finding was identified involving a loss of condenser vacuum caused by plant operators improperly removing a steam jet air ejector from service. Specifically, the licensee failed to isolate the steam jet air ejector from service as delineated in the system operating instructions. As a result, the plant experienced a loss of condenser vacuum and a plant transient. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-05676. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the initiating event and mitigating systems cornerstone. The inspectors performed a Phase 2 analysis using Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," of Manual Chapter 0609, "Significance Determination Process," and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors assumed that only the power conversion system was affected and all other mitigating systems were available. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding was related to the human performance crosscutting component of work practices in that the control room supervisor failed to ensure supervisory and management oversight of work activities such that nuclear safety is supported [H.4(c)] (Section 4OA2). Inspection Report# : 2007005 (pdf) Significance: Sep 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Reactor SCRAM due to Turbine Trip caused by Loss of Condenser Vacuum A self-revealing finding was identified involving the failure to properly calibrate the main condenser hydraulic vacuum switch that established a higher trip setpoint that would prematurely actuate an automatic turbine trip and reactor scram for a degraded main condenser vacuum condition. This issue was entered into the licensees corrective action program as condition Report CR-GGN-2007-02756. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the MC

0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance, because the finding did not contribute to the likelihood that mitigating equipment or functions would not be available following a reactor trip. The cause of the finding was related to the human performance crosscutting component of resources in that the calibration procedure did not provide clear instructions detailing the methodology to adjust the speed simulation screw to the required position.[H.2(c)] Inspection Report# : 2007004 (pdf) Mitigating Systems Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation (PMP) Door Seals Protecting Safety Related Equipment. The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. The inspectors identified that the door seals did not make contact with the door frame and the door had a significant amount of corrosion underneath the door seals, indicating long term degradation. The extent of condition review found three additional door seals with degraded conditions, including doors to the standby service water pump houses. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2008-01123 and 2008-01623. This finding was more than minor because the door seals represent a degrading condition that if left uncorrected could become a more significant safety concern. The inspectors determined this finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance since it did not represent an actual loss of safety function for the standby service water pumps or the diesel generators. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to properly identify the degraded conditions of the probable maximum precipitation door seals during their surveillance inspection. [P.1(a)] (Section 1R01) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement an Adequate Compensatory Fire Watch per Station Fire Protection Procedures. The inspectors identified a noncited violation of Facility Operating License Condition 2.C.41 for the failure to properly implement a compensatory fire watch per the station fire protection program. The inspectors performed a fire inspection of the auxiliary building electrical penetration room. The inspectors noted that plant personnel had not entered the room to perform a required fire watch. The inspectors questioned security personnel, reviewed the fire watch log and determined that the fire watch log had been checked off as completed. The completion time corresponded to the time the inspector was in the room. After further review and interviews with security personnel, the inspectors determined that the plant employee designated to perform the fire watch duties misunderstood the requirements for the fire watch. The employee had only verified the auxiliary building hallway area outside the room and failed to check inside the auxiliary building electrical penetration room as required. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-00869. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for outages or degradations. The inspectors assigned a high degradation rating due to the fact that automatic fire suppression system was tagged out and inoperable. Because the system was degraded without compensatory actions for approximately 2 hours, the inspectors used a duration factor of 0.01. The

inspectors used 2E-2 for a generic fire frequency area which corresponds to Table 1.4.2, Generic Fire Area Fire Frequencies for a switchgear room. The resulting calculated change in core damage frequency was 2E-4, which was greater than the high degradation Phase 1 Quantitative Screening Criteria of 1E-6, requiring a Phase 2 analysis. The inspectors consulted with a regional Senior Reactor Analyst and a simplified Phase 3 was performed using a duration factor of 2.3E-4 for the 2-hour time period, and the IPEEE specific room fire frequency of 7.2E-4. The resulting calculated change in core damage frequency was 1.7E-7, which would be less that the Phase 1 quantitative screening criteria. Using this information, the regional Senior Reactor Analyst, determined the finding to be of very low safety significance. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices in that the individual assigned to perform the fire watch proceeded in the face of uncertainty and failed to use appropriate human error prevention techniques. [H.4(a)] (Section 1R05) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure of Licensed Senior Reactor Operators to Maintain the Required Proficiency to Maintain Their License Current. The inspectors identified a noncited violation of 10 CFR 55.53.e, Conditions of License, for failure of licensed senior reactor operators to maintain the required proficiency to maintain their license current. Senior reactor operators standing the shift supervisor/shift technical advisor position were taking credit for senior reactor operator proficiency watches while standing this position. The normal shift complement of senior reactor operators consist of a shift manager, a control room supervisor, and a shift supervisor/shift technical advisor. When this issue was brought to the attention of operations management; they stopped the practice of the shift supervisor/shift technical advisor receiving senior reactor operator proficiency watch credit for standing that position. All shift supervisor/shift technical advisor senior reactor operators were inactivated. The plant issued a standing order that prohibited the shift supervisor/shift technical advisor to be allowed to perform the senior reactor operators oversight function in the control room and the shift manager or control room supervisor had to be in the control room at all times. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-01126. This finding was more than minor because if left uncorrected the finding could become a more significant safety concern. This finding affects the mitigating system cornerstone. The finding was determined to be of very low safety significance using the Licensed Operator Requalification Significance Determination Process since it related to operator license conditions and more than 20 percent of the affected individuals were deficient (Section 1R11). Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform a Required Technical Specification Surveillance. The inspectors identified a noncited violation of Technical Specifications 3.8.1, AC Sources-Operating, for the failure to perform a required surveillance following the loss of a required offsite power source. The plant suffered a loss of power from the Port Gibson 115 kV line during high winds. Due to the fact that there is no direct control room alarm to alert the operating crew, they were not immediately aware they had lost the offsite source of power. When the crew recognized the loss of the bus they only entered a potential limiting condition of operations, due to the crew failing to realize that this was one of the required offsite sources. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2008-00737 and 2008-01202. This finding was more than minor because it impacts the mitigating system cornerstone objective in that it affects the operability, availability, reliability of an offsite power source that supplies a bus that provides power to mitigating systems. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was of very low safety significance since it did not represent an actual loss of a safety function. The cause of this finding has a crosscutting aspect in the area of human performance associated with the resources attribute in that the operators did not have adequate procedural guidance to determine the loss of safety-related offsite power supply. [H.2(c)] (Section 1R22) Inspection Report# : 2008002 (pdf)

Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate Cracks in Standby Service Water Pump House Structure. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, for failing to implement effective corrective actions after identifying concrete cracking in the standby service water pump houses. The inspectors determined that the program that evaluates, monitors, and repairs cracks for all safety related structures only identified a single crack for the entire site and does not track other structural cracks previously identified in the corrective action program. The last program inspection had been performed as recently as October 25, 2007, and only identified the single crack that had been documented in previous inspections. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2007-05824. This finding was more than minor because the cracks represent a degrading condition that if left uncorrected could become more significant safety concern. The inspectors determined this finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was of very low safety significance since it did not represent an actual loss of a safety function. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee personnel failed to properly maintain and utilize the program for evaluating, tracking and repairing identified concrete cracks in safety related structures. [H.4(b)] (Section 4OA2) Inspection Report# : 2008002 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Venting Procedure for the Reactor Core Isolation Cooling System The inspectors identified a noncited violation of Criterion V, Instructions, Procedures, and Drawings, of 10 CFR Part 50, Appendix B for the failure to demonstrate compliance with Technical Specification Surveillance Requirement 3.5.3.1 due to an inadequate surveillance procedure. The reactor core isolation cooling system is vented at the injection valve through a hard-piped drain with no visual means of detecting air in the system. The inspectors determined that the procedure failed to contain adequate acceptance criteria to qualitatively or quantitatively assess abnormal amounts of air in the reactor core isolation cooling system. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-03818. The finding was greater than minor because it affects the procedure quality 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. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have a very low safety significance in that it did not result in the actual loss of the reactor core isolation cooling system, and was not potentially risk-significant due to external initiating events. Inspection Report# : 2007005 (pdf) Significance: Dec 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a root cause analysis for RHR heat exchanger B fouling, and implement corrective action to prevent recurrence A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for failure to perform an adequate cause analysis for fouling of the Residual Heat Removal Heat Exchanger B on the standby service water side, and implement corrective action to prevent recurrence. This fouling reduced the thermal performance margin to 0.6 percent, but was not treated as a significant condition adverse to quality within the corrective action program. The licensee chose to temporarily restore margin by increasing the flow rate, but this did not remove or stop the fouling from continuing to occur. This finding has cross cutting aspects in the decision-making area of Human Performance (H.1.b) because the licensees decision-making in response to this degraded condition did not use conservative criteria in deciding when to clean this heat exchanger, and did not verify that the underlying assumptions remained valid. Failure to treat Residual Heat Removal Heat Exchanger B degradation as a significant condition adverse to quality, and perform an adequate cause analysis, and implement corrective action to prevent recurrence was a performance deficiency. This was more than minor because, if left uncorrected, it could lead to a more significant safety concern in

that the system could become fouled enough to prevent removing the required heat load without the licensee recognizing this condition. This finding affected the Mitigating Systems and Barrier Integrity Cornerstones, since this component was required for both decay heat removal and containment heat removal functions. In accordance with the Phase 1 Significance Determination Process instructions, the significance was assessed using the Mitigating Systems Cornerstone, since this represented the dominant risk. This finding was determined to have very low safety significance (Green) during a Phase 1 Significance Determination Process, since it was confirmed to not involve loss of the design heat removal capability. This issue was entered into the licensees corrective action program under Condition Report 2007-5766. (Section 4OA2.e.1(b)(1)) Inspection Report# : 2007008 (pdf) Significance: Dec 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate thermal performance testing of the residual heat removal heat exchangers A noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, was identified because the licensees thermal performance test procedures for the residual heat removal heat exchangers were inadequate to ensure the quality of the test results. Specifically, the test procedure failed to specify adequate prerequisites for minimum heat load and use of high-accuracy instrumentation. This resulted in test results used to meet commitments for the Generic Letter 89-13 test program which provided little useful information due to high inaccuracy. Failure to adequately test and trend the thermal performance of the residual heat removal heat exchangers was a performance deficiency because it masked the actual thermal performance to the point where the licensee did not recognize the onset of fouling. The team determined that these heat exchangers began to experience fouling between 1997 and 1998, but this was not recognized. In the case of Residual Heat Removal Heat Exchanger B, the degraded performance was determined to be sufficient to make the fouling factor exceed the design value, necessitating compensatory action to be able to show continued operability. This was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the system could become fouled enough to prevent removing the required heat load without the licensee recognizing this condition. This finding affected the Mitigating Systems and Barrier Integrity Cornerstones, since this component was required for both decay heat removal and containment heat removal functions. In accordance with the Phase 1 SDP instructions, the significance was assessed using the Mitigating Systems Cornerstone, since this represented the dominant risk. This finding was determined to have very low safety significance (Green) during a Phase 1 Significance Determination Process, since it was confirmed to not involve loss of the design heat removal capability. This issue was entered into the licensees corrective action program under Condition Report 2008-0412. (Section 1R07) Inspection Report# : 2007008 (pdf) Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of the Control Rod Drive System. The inspectors identified a Green noncited violation (NCV) of 10 CFR Part 50.65(a)(2) for the failure to adequately monitor the performance of the control rod drive system. Specifically, the licensee failed to adequately perform a functional failure determination for a degraded flow control valve. Following licensee review of this condition the system was placed in the maintenance rule (a)(1) monitoring status. This finding was more than minor since the degraded control rod drive flow control valve caused the system to be placed in the (a)(1) monitoring status. This finding was characterized under the significance determination process as having a very low safety significance, because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system, nor did it cause a component to become inoperable. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making, because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions. H.1(b) Inspection Report# : 2007004 (pdf) Significance: Sep 30, 2007

Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish a Formal Procedure to Monitor Outdoor Air Temperatures. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to translate a design basis limit for outdoor air temperature into an instruction or procedure. The licensee established a new Updated Final Safety Analysis Report maximum outdoor air temperature of 102.5 degrees F. If outside air temperatures exceeded 102.5 degrees F, safety-related equipment, which are located in rooms that are cooled by outdoor air (i.e., standby service water pump room), would be operationally challenged. The inspectors identified that no instruction or procedure was established to monitor high outside temperature or subsequent actions established in the event the design basis temperature limit is exceeded. The inspectors determined that the finding was more than minor because the finding affects the mitigating system cornerstone objective of ensuring the reliability of the standby service water system that responds to initiating events to prevent undesirable conditions. Using the Phase 1 worksheet in Inspection Manual Chapter 0609, "Significance Determination Process," this finding is determined to be of very low safety significance because there was no actual loss of a safety function, and the design basis limits had not been exceeded. The inspectors determined that the finding has a crosscutting aspect in the area of human performance decision making because the licensee failed to use conservative assumptions in determining not to establish a procedure or instruction to monitor high outside temperature for design limits on the standby service water pump room. H.1(b) Inspection Report# : 2007004 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Performance of the Leakage Detection System The inspectors identified a Green noncited violation involving the failure to adequately monitor the performance of the leakage detection system in accordance with 10CFR50.65(a)(2). Specifically, the licensee failed to account for the functional failure of a temperature switch which resulted in exceeding the performance criteria for the leakage detection system. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-2955. This finding was greater than minor since violations of 10 CFR Part 50.65(a)(2) necessarily involve degraded system performance which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the maintenance rule aspect of the finding did not lead to an actual loss of safety function of the system nor did it cause a component to be inoperable. This finding has a crosscutting aspect in the area of human performance associated with work practices in that the licensee failed to use human error prevention techniques such as self checking and peer checking when utilizing the maintenance rule database (H.4(a)). Inspection Report# : 2007003 (pdf) Significance: Jun 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Foreign Material Controls During Reactor Feed Pump Maintenance A self-revealing Green finding was identified for inadequate foreign material controls during maintenance. Specifically, a foreign material exclusion device was left inside the reactor feed Pump B lube oil system following maintenance activities, which prevented placing the pump in service during reactor startup. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-2158. The finding was more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. The inspectors determined this finding required a Phase 2 analysis because it resulted in the loss of function of a single train of the power conversion system (reactor feed) for greater than 24 hours. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance because of the availability of the condensate booster pumps and emergency core cooling systems. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources because licensee personnel were not adequately trained to consistently implement the foreign material exclusion program (H.2(b)). Inspection Report# : 2007003 (pdf)

Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure for Safety-Related Breaker Inspections The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure to identify loose and missing fasteners on the standby service water Train B bus feeder breaker. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-3081. This finding was more than minor because the failure to ensure that loose parts are not present in safety related breakers, if left uncorrected, could become a more significant safety concern. Using the Significance Determination Process Phase 1 Screening Worksheet in Appendix A of Inspection Manual Chapter 0609, the inspectors determined the finding was of very low safety significance because it did not result in a loss of operability. Inspection Report# : 2007003 (pdf) Barrier Integrity Significance: Mar 22, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Design Control of HPCS Minimum Flow Valve Motor-Operated Valve Over Current Setpoint. The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to properly set the over current trip setpoint for the high pressure core spray minimum flow motor operated valve. This resulted in a spurious over current trip of the valve breaker during a high pressure core spray momentary pump start for breaker operability following post Division 3 emergency core cooling system testing. As a result of the trip, the high pressure core spray minimum flow valve failed open. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-01201. The finding was more than minor because it was associated with the barrier integrity cornerstone to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the MC 0609, "Significance Determination Process," Phase 1 worksheet, the finding was determined to have very low safety significance since it did not result in a loss of the containment barrier. Additionally, the issue was screened and determined to not impact the High Pressure Core Spray mitigating system function. (Section 4OA3) Inspection Report# : 2008002 (pdf) Significance: Sep 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures Caused Loss of Decay Heat Removal in the Spent Fuel Pool. A self-revealing Green non-cited violation of Technical Specifications 5.4.1(a) was identified involving the failure to adequately follow procedure to align valves in the fuel pool cooling and cleanup system. The valves were aligned in the wrong sequence, contrary to the system operating instructions, causing both fuel pool cooling and cleanup pumps to trip and a subsequent loss of fuel pool cooling. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-04284. The finding is more than minor, since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has a very low safety significance since it only represents a degradation of the radiological barrier function provided by the spent fuel pool system. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel failed to follow the correct sequence of valve manipulations required by procedure. [H.4 (b)] Inspection Report# : 2007004 (pdf)

Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Repair Crack in Containment Building Structure The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving the failure to correct a crack in the ceiling of the reactor water cleanup heat exchanger room internal to the containment building structure. Specifically, the licensee identified the crack in 1987 but failed to complete planned corrective actions to evaluate or repair the crack during Refueling Outage 2. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-1970. This finding was more than minor because the reactor water cleanup (RWCU) ceiling crack represented a degrading condition that if left uncorrected could become a more significant safety concern. The inspectors determined this finding affected the Barrier Integrity cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not represent an actual open pathway in the physical integrity of the reactor containment or an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. Inspection Report# : 2007003 (pdf) Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Procedure The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for the failure to provide a detailed work order package to perform vent and fill operations on a pressure transmitter. Specifically, the licensee did not provide appropriate instructions in a work order package to properly isolate pressure Transmitter 1N64N006B prior to opening the drain valve. Consequently, this resulted in the release of radioactive gas from the system and an unplanned and unintended exposure for two individuals involved in the work activity. The finding is more than minor because it is associated with the occupational radiation safety attribute of program and process and affected the cornerstone objective because it involved unplanned and unintended dose to two workers. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because: it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a cross-cutting aspect in the area of work control associated with work planning because the licensee failed to properly plan work activities by incorporating specific plant system details into the work order to allow the instrumentation and control technicians to properly drain a pressure transmitter [H.3(a)] (Section 2OS2). Inspection Report# : 2007005 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Radiological Hazard Caused by Foreign Material Retrieval from the Reactor Vessel The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 20.1501(a) because the licensee failed to evaluate the radiological hazard of foreign material retrieval from the reactor vessel. A contract radiation protection technician misinterpreted his survey instrument readings, picked up a bolt with a radiation dose rate of 19.9 rem per hour, and received a shallow dose equivalent of 41 millirems. The radiation protection technician was alerted to the problem by an electronic dosimeter alarm. As corrective action, the licensee revised the appropriate radiation work permit template to incorporate a dose rate limit for items removed from pools and included a discussion of the

violation in radiation protection training. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the lack of knowledge of radiological conditions could increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable (ALARA) planning or work control issue; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work practices because the workers failed to use error prevention techniques such as peer checking and self checking (H.4(a)). Inspection Report# : 2007003 (pdf) Significance: Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post and Control a High Radiation Area The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.7.1 resulting from a failure to post and control a high radiation area. Room 0R123 on the 93-foot elevation of the radwaste building had dose rates as high as 265 millirems per hour at 30 centimeters from the G17D069 filter housing and was not posted and controlled as a high radiation area. The licensee was alerted to the situation when the electronic dosimeters of two radwaste operators alarmed when they entered the higher dose rates. Poor communications between operations and radiation protection personnel contributed to the failure to identify the high radiation area. Radiation protection supervisors stated they were unaware at the time of the operators dose rate alarms that reactor water cleanup reject flow was approximately twice the normal flow rate and both of the reactor water cleanup demineralizers had been out of service from approximately 3:00 p.m. on May 19 until 9:00 a.m. on May 20, 2007. As immediate corrective action, the area was barricaded and conspicuously posted as a high radiation area. Additional planned corrective actions were still being evaluated. This finding is greater than minor because it is associated with the occupational radiation safety program and process attribute and affected the cornerstone objective, in that the failure to post and control a high radiation area had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance because it did not involve: (1) an as low as is reasonably achievable (ALARA) planning or work control issue; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. Additionally, this finding has a crosscutting aspect in the area of human performance associated with work control because the licensee failed to ensure proper communication, coordination, and cooperation during activities in which interdepartmental coordination was necessary to assure plant and human performance (H.3(b)). Inspection Report# : 2007003 (pdf) Public Radiation Safety Significance: Sep 14, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Function-Specific Training to Hazardous Material Workers. The team identified a noncited violation (NCV) of 10 CFR 71.5 because the licensee failed to provide required training to hazardous material workers involved in the shipment of radioactive material. Specifically, the licensee did not provide function-specific training, pursuant to 49 CFR 172.704(a) of Department of Transportation shipping regulations, to maintenance personnel involved in the reassembly the shipping casks. Corrective actions are still being evaluated; however, the licensee plans to provide hazardous material training to these employees. The licensee documented this issue in the corrective action program as Condition Report-GGN-2007-04572. The finding is greater than minor because it is associated with the Public Radiation Safety Cornerstone attribute of program and process and affects the cornerstone objective. Inadequate training of hazardous material workers regarding the reassembly and loading of shipping casks has a potential impact on public dose and on the licensees ability to safely package and transport radioactive material on public roadways. The violation involved an occurrence in the licensees radioactive material transportation program that is contrary to NRC or Department of Transportation

regulations. When processed through the Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it: (1) was associated with radioactive material control, (2) involved the licensees program for radioactive material packaging and transportation, (3) did not cause radiation limits to be exceeded, (4) did not result in a breach of package during transit, (5) did not involve a certificate of compliance issue, (6) did not involve a non-compliance with low level burial ground, and (7) did not involve a failure to make notifications or to provide emergency information. In addition, this finding had cross-cutting aspects in the area of human performance in the component of resources because the licensee did not ensure the availability and adequacy of training for hazardous material workers involved in the shipment of radioactive material was available. Hazardous material workers are required to be trained and qualified to prepare radioactive material shipments for transport. (H.2.b) (Section 2PS2) Inspection Report# : 2007007 (pdf) Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: N/A Dec 30, 2007 Identified By: NRC Item Type: FIN Finding Identification and resolution of problems The inspectors reviewed approximately 200 condition reports, work orders, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities. The team concluded that the licensee was generally effective in identifying, evaluating, and correcting problems. Corrective actions, when specified, were generally implemented in a timely manner, although the team identified a significant number of longstanding equipment problems that were not being resolved in a timely manner. The team concluded that the licensee continued to have problems with the quality of operability assessments, and this was not being effectively addressed. The licensee performed quality higher-tier self-assessments, but the overall effectiveness was reduced by being slow to implement recommended improvements. The team concluded that the licensee was making progress in their efforts to address a trend in human performance, but this has not yet been completely effective. On the basis of 32 interviews conducted during this inspection, workers at the site felt free to report problems to their management, and were willing to use the corrective action program. An increased awareness and confidence in the Employee Concerns Program was also apparent. The team concluded that a positive safety-conscious work environment exists at Grand Gulf Nuclear Station. Inspection Report# : 2007008 (pdf) Last modified : June 05, 2008

Grand Gulf 1 2Q/2008 Plant Inspection Findings Initiating Events Significance: Mar 22, 2008 Identified By: NRC Item Type: FIN Finding Ineffective Corrective Actions in Response to Resin in the Electro-hydraulic Control System. The inspectors identified a finding involving ineffective corrective actions in response to resin intrusion in the electro-hydraulic control system. The inspectors reviewed the corrective actions from a condition report involving a resin intrusion into the electro-hydraulic control system via a failed temporary ion-exchange filter in 2003. Review of the corrective actions associated with the 2003 event revealed that a long-range recovery plan was developed to remove resin from the electro-hydraulic control system. However, the recovery plan corrective actions were closed without licensee actions to remove resin from the electro-hydraulic control system. The failure to implement effective corrective actions following the 2003 resin intrusion event directly resulted in electro-hydraulic control stability issues seen in the fall of 2007, including reactor pressure perturbations and reductions in reactor power. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2007-04972. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 worksheet, the finding was determined to have very low safety significance because the finding did not contribute to the likelihood that mitigating equipment would not be available following a reactor trip. (Section 4OA2) Inspection Report# : 2008002 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Improper Control of Troubleshooting Causes a Loss of Condenser Vacuum The inspectors identified a finding involving a loss of condenser vacuum caused by improper troubleshooting of the seal steam pressure controller. Specifically, the licensee failed to provide adequate work instructions and procedural limitations during troubleshooting of the seal steam pressure controller. As a result, the plant experienced a loss of condenser vacuum and a plant transient. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-04626. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the initiating event and mitigating systems cornerstone. The inspectors performed a Phase 2 analysis using Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," of Manual Chapter 0609, "Significance Determination Process," and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors assumed that only the power conversion system was affected and all other mitigating systems were available. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding was related to the human performance crosscutting component of decision making, in that the licensee failed to use conservative assumptions during troubleshooting activities and performed these activities without determining the validity of the troubleshooting instructions and identifying possible unintended consequences [H.1(b)] (Section 4OA3). Inspection Report# : 2007005 (pdf)

Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Engineering Review of Plant Service Water Modification A self-revealing finding was identified involving the failure of a plant service water piping flange due to an improper flow control valve design modification. Specifically, the licensee failed to perform an adequate review of an engineering modification and the maintenance work orders did not have detailed installation instructions. As a result, the plant experienced a plant transient. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-05040. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because the finding did not contribute to the likelihood that mitigating equipment or functions would not be available following a reactor trip. The cause of the finding was related to the human performance crosscutting component of work practices in that the responsible engineers failed to perform adequate self and peer checking during the development and review of the design modification to the plant service water flow control check valves [H.4(a)] (Section 4OA3). Inspection Report# : 2007005 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Follow Procedure Results in Loss of Condener Vacuum A self-revealing finding was identified involving a loss of condenser vacuum caused by plant operators improperly removing a steam jet air ejector from service. Specifically, the licensee failed to isolate the steam jet air ejector from service as delineated in the system operating instructions. As a result, the plant experienced a loss of condenser vacuum and a plant transient. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-05676. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the initiating event and mitigating systems cornerstone. The inspectors performed a Phase 2 analysis using Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," of Manual Chapter 0609, "Significance Determination Process," and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors assumed that only the power conversion system was affected and all other mitigating systems were available. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding was related to the human performance crosscutting component of work practices in that the control room supervisor failed to ensure supervisory and management oversight of work activities such that nuclear safety is supported [H.4(c)] (Section 4OA2). Inspection Report# : 2007005 (pdf) Significance: Sep 30, 2007 Identified By: Self-Revealing Item Type: FIN Finding Reactor SCRAM due to Turbine Trip caused by Loss of Condenser Vacuum A self-revealing finding was identified involving the failure to properly calibrate the main condenser hydraulic vacuum switch that established a higher trip setpoint that would prematurely actuate an automatic turbine trip and reactor scram for a degraded main condenser vacuum condition. This issue was entered into the licensees corrective action program as condition Report CR-GGN-2007-02756. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using the MC

0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to have very low safety significance, because the finding did not contribute to the likelihood that mitigating equipment or functions would not be available following a reactor trip. The cause of the finding was related to the human performance crosscutting component of resources in that the calibration procedure did not provide clear instructions detailing the methodology to adjust the speed simulation screw to the required position.[H.2(c)] Inspection Report# : 2007004 (pdf) Mitigating Systems Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate Fire Wrap Testing Discrepancies A noncited violation of License Condition 2.C(41), "Fire Protection Program," was identified because the licensee failed to evaluate vendor fire test results to ensure that deviations from the acceptance criteria did not result in a reduction in the effectiveness of the approved Fire Protection Program. The licensee replaced existing fire barrier material installed on conduits with 3M Interam fire wrap without recognizing that applicable NRC test criteria were not met. As a result, the licensee failed to perform an evaluation to determine whether the test results would result in a reduction in the effectiveness of the fire protection provided to the cables inside the affected conduits. The new fire wrap was installed to protect redundant trains of cables necessary for safe shutdown between 2004 and 2007. This finding was entered into the licensees corrective action program under Condition Report 2008-01910. The licensee took prompt compensatory measures and implemented hourly fire watches while the issue was being evaluated. Failure to properly evaluate vendor fire test results that did not satisfy the acceptance criteria in Generic Letter 86-10, Supplement 1 prior to changing the existing fire wrap with 3M Interam fire wrap as required by the approved Fire Protection Program was a performance deficiency. This finding was more than minor because it affected the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone Objective to ensure the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. This performance deficiency was also similar to the "more than minor" portion of Inspection Manual Chapter 0612, Appendix B, Example 3.i, in that an engineering evaluation was necessary to determine the acceptability of the existing fire wrap to perform its intended function. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected fire protection defense-in-depth strategies involving post-fire safe shutdown systems. This finding screened as having very low safety significance because it involved a fire barrier with a low degradation, since the nonconforming condition was subsequently determined to provide an acceptable margin to damage for the cables being protected. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Fire Brigade Members Assigned Responsibilities That Conflicted with Fire Brigade Responsibilities. A noncited violation of License Condition 2.C.(41) was identified for failure to maintain required staffing available to respond to a fire. Specifically, the approved Fire Protection Program requires that a five-person fire brigade be available onsite at all times and not assigned duties that conflict with the duties of the fire brigade. Contrary to this, on three occasions in March 2008, operators assigned as fire brigade members were directed to perform operator rounds at the radial wells. Because the Mississippi River was at flood stage, this required traveling by boat, so the operators were unable to return to the plant promptly for approximately 2 hours. This was further complicated by the fact that operator/fire brigade radios did not work during most of the boat trip and in the vicinity of the most distant well, meaning that operators could not be quickly recalled. This finding was entered into the corrective action program under Condition Report 2008-01616. This finding had a crosscutting aspect in the area of Human Performance - Work Control (H.3.b) because the licensee did not ensure that different job activities were coordinated to ensure that the fire brigade remained available at all times. Failure to maintain a fully staffed fire brigade available onsite at all times was a performance deficiency. This finding was more than minor because it affected the protection from external factors (fire) attribute of the Mitigating Systems

Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesireable consequences. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected a fire protection defense-in-depth element. This finding was assigned a low degradation rating because the operations shift during the times when the fire brigade member was unavailable included extra fire brigade-trained personnel that could supplement the fire brigade. The delay in a replacement person reporting to the scene of a fire would not have impacted the initial fire fighting effort, since enough fire brigade personnel were available to perform the functions. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Plant Modification Blocked Access for Manual Firefighting. A noncited violation of License Condition 2.C.(41), "Fire Protection Program," was identified related to making a plant change that negatively impacted the effectiveness of the approved Fire Protection Program. The team identified that the licensee had permanently blocked the door to the abandoned Unit 2 portion of the joint control room without performing a fire protection impact evaluation. The only remaining access path was a small hatch that would have made it difficult for fire fighters to gain access with protective clothing and equipment. A fire in this area could threaten operations in the Unit 1 control room if not promptly suppressed. This finding was determined to have a cross-cutting aspect in problem identification and resolution timeliness (P.1.d) because fire protection personnel recognized that a new access door was needed in 2006, but no substantial action had been taken to install it by the time of this inspection. This finding was entered in to the licensees corrective action program under Condition Reports 2008-001893 and 2008-01913. Blocking access to the Unit 2 control room area and not promptly restoring access to allow manual fire suppression was a performance deficiency. This finding was more than minor because it affected the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected a fire protection defense-in-depth element. This finding was determined to have very low safety significance because all potential fire ignition sources in the affected area screened out in Task 2.3.4 in the Phase 2 evaluation. There were no ignition sources because the licensee had removed electrical power from this area, and administratively prevented hot work and storage of transient combustible material from this area. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure That Potential Damage to Motor-Operated Valve Circuits Would Not Prevent Safe Shutdown. A noncited violation of 10 CFR Part 50, Appendix R, Section III.G.1.a was identified because the licensee failed to evaluate the impact of a potential motor operated valve failure mechanism on the ability to implement post-fire safe shutdown following a control room evacuation. The team identified that the Residual Heat Removal Pump Minimum Flow Valve F064A could be damaged by fire in the control room and not be available to perform its safe shutdown function. This finding involved mechanistic damage due to hot shorts as described in Information Notice 92-18, "Potential for Loss of Remote Shutdown Capability During Control Room Fire." The licensee had incorrectly interpreted this operating experience and concluded that no action was required. This finding was entered into the corrective action program under Condition Reports 1999-0236 and 2008-01904. The team determined that failure to ensure that components necessary to safely shutdown the reactor would remain operable following a fire was a performance deficiency. This deficiency was more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to external events (fire) to prevent undesirable consequences. The Phase 3 risk evaluation performed by the senior reactor analyst determined this deficiency had very low safety significance because the probability of having a fire in either of the two control room panels where the postulated damage could occur and lead to a control room evacuation was very low.

Inspection Report# : 2008006 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation (PMP) Door Seals Protecting Safety Related Equipment. The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. The inspectors identified that the door seals did not make contact with the door frame and the door had a significant amount of corrosion underneath the door seals, indicating long term degradation. The extent of condition review found three additional door seals with degraded conditions, including doors to the standby service water pump houses. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2008-01123 and 2008-01623. This finding was more than minor because the door seals represent a degrading condition that if left uncorrected could become a more significant safety concern. The inspectors determined this finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance since it did not represent an actual loss of safety function for the standby service water pumps or the diesel generators. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to properly identify the degraded conditions of the probable maximum precipitation door seals during their surveillance inspection. [P.1(a)] (Section 1R01) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement an Adequate Compensatory Fire Watch per Station Fire Protection Procedures. The inspectors identified a noncited violation of Facility Operating License Condition 2.C.41 for the failure to properly implement a compensatory fire watch per the station fire protection program. The inspectors performed a fire inspection of the auxiliary building electrical penetration room. The inspectors noted that plant personnel had not entered the room to perform a required fire watch. The inspectors questioned security personnel, reviewed the fire watch log and determined that the fire watch log had been checked off as completed. The completion time corresponded to the time the inspector was in the room. After further review and interviews with security personnel, the inspectors determined that the plant employee designated to perform the fire watch duties misunderstood the requirements for the fire watch. The employee had only verified the auxiliary building hallway area outside the room and failed to check inside the auxiliary building electrical penetration room as required. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-00869. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for outages or degradations. The inspectors assigned a high degradation rating due to the fact that automatic fire suppression system was tagged out and inoperable. Because the system was degraded without compensatory actions for approximately 2 hours, the inspectors used a duration factor of 0.01. The inspectors used 2E-2 for a generic fire frequency area which corresponds to Table 1.4.2, Generic Fire Area Fire Frequencies for a switchgear room. The resulting calculated change in core damage frequency was 2E-4, which was greater than the high degradation Phase 1 Quantitative Screening Criteria of 1E-6, requiring a Phase 2 analysis. The inspectors consulted with a regional Senior Reactor Analyst and a simplified Phase 3 was performed using a duration factor of 2.3E-4 for the 2-hour time period, and the IPEEE specific room fire frequency of 7.2E-4. The resulting calculated change in core damage frequency was 1.7E-7, which would be less that the Phase 1 quantitative screening criteria. Using this information, the regional Senior Reactor Analyst, determined the finding to be of very low safety significance. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices in that the individual assigned to perform the fire watch proceeded in the face of uncertainty and failed to use appropriate human error prevention techniques. [H.4(a)] (Section 1R05)

Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure of Licensed Senior Reactor Operators to Maintain the Required Proficiency to Maintain Their License Current. The inspectors identified a noncited violation of 10 CFR 55.53.e, Conditions of License, for failure of licensed senior reactor operators to maintain the required proficiency to maintain their license current. Senior reactor operators standing the shift supervisor/shift technical advisor position were taking credit for senior reactor operator proficiency watches while standing this position. The normal shift complement of senior reactor operators consist of a shift manager, a control room supervisor, and a shift supervisor/shift technical advisor. When this issue was brought to the attention of operations management; they stopped the practice of the shift supervisor/shift technical advisor receiving senior reactor operator proficiency watch credit for standing that position. All shift supervisor/shift technical advisor senior reactor operators were inactivated. The plant issued a standing order that prohibited the shift supervisor/shift technical advisor to be allowed to perform the senior reactor operators oversight function in the control room and the shift manager or control room supervisor had to be in the control room at all times. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-01126. This finding was more than minor because if left uncorrected the finding could become a more significant safety concern. This finding affects the mitigating system cornerstone. The finding was determined to be of very low safety significance using the Licensed Operator Requalification Significance Determination Process since it related to operator license conditions and more than 20 percent of the affected individuals were deficient (Section 1R11). Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform a Required Technical Specification Surveillance. The inspectors identified a noncited violation of Technical Specifications 3.8.1, AC Sources-Operating, for the failure to perform a required surveillance following the loss of a required offsite power source. The plant suffered a loss of power from the Port Gibson 115 kV line during high winds. Due to the fact that there is no direct control room alarm to alert the operating crew, they were not immediately aware they had lost the offsite source of power. When the crew recognized the loss of the bus they only entered a potential limiting condition of operations, due to the crew failing to realize that this was one of the required offsite sources. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2008-00737 and 2008-01202. This finding was more than minor because it impacts the mitigating system cornerstone objective in that it affects the operability, availability, reliability of an offsite power source that supplies a bus that provides power to mitigating systems. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was of very low safety significance since it did not represent an actual loss of a safety function. The cause of this finding has a crosscutting aspect in the area of human performance associated with the resources attribute in that the operators did not have adequate procedural guidance to determine the loss of safety-related offsite power supply. [H.2(c)] (Section 1R22) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate Cracks in Standby Service Water Pump House Structure. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, for failing to implement effective corrective actions after identifying concrete cracking in the standby service water pump houses. The inspectors determined that the program that evaluates, monitors, and repairs cracks for all safety related structures only identified a single crack for the entire site and does not track other structural cracks previously identified in the corrective action program. The last program inspection had been performed as recently as October 25, 2007, and only

identified the single crack that had been documented in previous inspections. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2007-05824. This finding was more than minor because the cracks represent a degrading condition that if left uncorrected could become more significant safety concern. The inspectors determined this finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was of very low safety significance since it did not represent an actual loss of a safety function. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee personnel failed to properly maintain and utilize the program for evaluating, tracking and repairing identified concrete cracks in safety related structures. [H.4(b)] (Section 4OA2) Inspection Report# : 2008002 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Venting Procedure for the Reactor Core Isolation Cooling System The inspectors identified a noncited violation of Criterion V, Instructions, Procedures, and Drawings, of 10 CFR Part 50, Appendix B for the failure to demonstrate compliance with Technical Specification Surveillance Requirement 3.5.3.1 due to an inadequate surveillance procedure. The reactor core isolation cooling system is vented at the injection valve through a hard-piped drain with no visual means of detecting air in the system. The inspectors determined that the procedure failed to contain adequate acceptance criteria to qualitatively or quantitatively assess abnormal amounts of air in the reactor core isolation cooling system. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-03818. The finding was greater than minor because it affects the procedure quality 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. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have a very low safety significance in that it did not result in the actual loss of the reactor core isolation cooling system, and was not potentially risk-significant due to external initiating events. Inspection Report# : 2007005 (pdf) Significance: Dec 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a root cause analysis for RHR heat exchanger B fouling, and implement corrective action to prevent recurrence A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for failure to perform an adequate cause analysis for fouling of the Residual Heat Removal Heat Exchanger B on the standby service water side, and implement corrective action to prevent recurrence. This fouling reduced the thermal performance margin to 0.6 percent, but was not treated as a significant condition adverse to quality within the corrective action program. The licensee chose to temporarily restore margin by increasing the flow rate, but this did not remove or stop the fouling from continuing to occur. This finding has cross cutting aspects in the decision-making area of Human Performance (H.1.b) because the licensees decision-making in response to this degraded condition did not use conservative criteria in deciding when to clean this heat exchanger, and did not verify that the underlying assumptions remained valid. Failure to treat Residual Heat Removal Heat Exchanger B degradation as a significant condition adverse to quality, and perform an adequate cause analysis, and implement corrective action to prevent recurrence was a performance deficiency. This was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the system could become fouled enough to prevent removing the required heat load without the licensee recognizing this condition. This finding affected the Mitigating Systems and Barrier Integrity Cornerstones, since this component was required for both decay heat removal and containment heat removal functions. In accordance with the Phase 1 Significance Determination Process instructions, the significance was assessed using the Mitigating Systems Cornerstone, since this represented the dominant risk. This finding was determined to have very low safety significance (Green) during a Phase 1 Significance Determination Process, since it was confirmed to not involve loss of the design heat removal capability. This issue was entered into the licensees corrective action program under Condition Report 2007-5766. (Section 4OA2.e.1(b)(1)) Inspection Report# : 2007008 (pdf)

Significance: Dec 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate thermal performance testing of the residual heat removal heat exchangers A noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, was identified because the licensees thermal performance test procedures for the residual heat removal heat exchangers were inadequate to ensure the quality of the test results. Specifically, the test procedure failed to specify adequate prerequisites for minimum heat load and use of high-accuracy instrumentation. This resulted in test results used to meet commitments for the Generic Letter 89-13 test program which provided little useful information due to high inaccuracy. Failure to adequately test and trend the thermal performance of the residual heat removal heat exchangers was a performance deficiency because it masked the actual thermal performance to the point where the licensee did not recognize the onset of fouling. The team determined that these heat exchangers began to experience fouling between 1997 and 1998, but this was not recognized. In the case of Residual Heat Removal Heat Exchanger B, the degraded performance was determined to be sufficient to make the fouling factor exceed the design value, necessitating compensatory action to be able to show continued operability. This was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the system could become fouled enough to prevent removing the required heat load without the licensee recognizing this condition. This finding affected the Mitigating Systems and Barrier Integrity Cornerstones, since this component was required for both decay heat removal and containment heat removal functions. In accordance with the Phase 1 SDP instructions, the significance was assessed using the Mitigating Systems Cornerstone, since this represented the dominant risk. This finding was determined to have very low safety significance (Green) during a Phase 1 Significance Determination Process, since it was confirmed to not involve loss of the design heat removal capability. This issue was entered into the licensees corrective action program under Condition Report 2008-0412. (Section 1R07) Inspection Report# : 2007008 (pdf) Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of the Control Rod Drive System. The inspectors identified a Green noncited violation (NCV) of 10 CFR Part 50.65(a)(2) for the failure to adequately monitor the performance of the control rod drive system. Specifically, the licensee failed to adequately perform a functional failure determination for a degraded flow control valve. Following licensee review of this condition the system was placed in the maintenance rule (a)(1) monitoring status. This finding was more than minor since the degraded control rod drive flow control valve caused the system to be placed in the (a)(1) monitoring status. This finding was characterized under the significance determination process as having a very low safety significance, because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system, nor did it cause a component to become inoperable. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making, because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions. H.1(b) Inspection Report# : 2007004 (pdf) Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish a Formal Procedure to Monitor Outdoor Air Temperatures. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to translate a design basis limit for outdoor air temperature into an instruction or procedure. The licensee established a new Updated Final Safety Analysis Report maximum outdoor air temperature of 102.5 degrees F. If outside air temperatures exceeded 102.5 degrees F, safety-related equipment, which are located in rooms that are cooled by outdoor air (i.e., standby service water pump room), would be operationally challenged. The inspectors identified that no instruction or procedure was established to monitor high outside temperature or subsequent actions established in the event the design basis temperature limit is exceeded.

The inspectors determined that the finding was more than minor because the finding affects the mitigating system cornerstone objective of ensuring the reliability of the standby service water system that responds to initiating events to prevent undesirable conditions. Using the Phase 1 worksheet in Inspection Manual Chapter 0609, "Significance Determination Process," this finding is determined to be of very low safety significance because there was no actual loss of a safety function, and the design basis limits had not been exceeded. The inspectors determined that the finding has a crosscutting aspect in the area of human performance decision making because the licensee failed to use conservative assumptions in determining not to establish a procedure or instruction to monitor high outside temperature for design limits on the standby service water pump room. H.1(b) Inspection Report# : 2007004 (pdf) Barrier Integrity Significance: Mar 22, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Design Control of HPCS Minimum Flow Valve Motor-Operated Valve Over Current Setpoint. The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to properly set the over current trip setpoint for the high pressure core spray minimum flow motor operated valve. This resulted in a spurious over current trip of the valve breaker during a high pressure core spray momentary pump start for breaker operability following post Division 3 emergency core cooling system testing. As a result of the trip, the high pressure core spray minimum flow valve failed open. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-01201. The finding was more than minor because it was associated with the barrier integrity cornerstone to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the MC 0609, "Significance Determination Process," Phase 1 worksheet, the finding was determined to have very low safety significance since it did not result in a loss of the containment barrier. Additionally, the issue was screened and determined to not impact the High Pressure Core Spray mitigating system function. (Section 4OA3) Inspection Report# : 2008002 (pdf) Significance: Sep 30, 2007 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedures Caused Loss of Decay Heat Removal in the Spent Fuel Pool. A self-revealing Green non-cited violation of Technical Specifications 5.4.1(a) was identified involving the failure to adequately follow procedure to align valves in the fuel pool cooling and cleanup system. The valves were aligned in the wrong sequence, contrary to the system operating instructions, causing both fuel pool cooling and cleanup pumps to trip and a subsequent loss of fuel pool cooling. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2007-04284. The finding is more than minor, since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has a very low safety significance since it only represents a degradation of the radiological barrier function provided by the spent fuel pool system. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel failed to follow the correct sequence of valve manipulations required by procedure. [H.4 (b)] Inspection Report# : 2007004 (pdf) Emergency Preparedness

Occupational Radiation Safety Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Procedure The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for the failure to provide a detailed work order package to perform vent and fill operations on a pressure transmitter. Specifically, the licensee did not provide appropriate instructions in a work order package to properly isolate pressure Transmitter 1N64N006B prior to opening the drain valve. Consequently, this resulted in the release of radioactive gas from the system and an unplanned and unintended exposure for two individuals involved in the work activity. The finding is more than minor because it is associated with the occupational radiation safety attribute of program and process and affected the cornerstone objective because it involved unplanned and unintended dose to two workers. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because: it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a cross-cutting aspect in the area of work control associated with work planning because the licensee failed to properly plan work activities by incorporating specific plant system details into the work order to allow the instrumentation and control technicians to properly drain a pressure transmitter [H.3(a)] (Section 2OS2). Inspection Report# : 2007005 (pdf) Public Radiation Safety Significance: Sep 14, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Function-Specific Training to Hazardous Material Workers. The team identified a noncited violation (NCV) of 10 CFR 71.5 because the licensee failed to provide required training to hazardous material workers involved in the shipment of radioactive material. Specifically, the licensee did not provide function-specific training, pursuant to 49 CFR 172.704(a) of Department of Transportation shipping regulations, to maintenance personnel involved in the reassembly the shipping casks. Corrective actions are still being evaluated; however, the licensee plans to provide hazardous material training to these employees. The licensee documented this issue in the corrective action program as Condition Report-GGN-2007-04572. The finding is greater than minor because it is associated with the Public Radiation Safety Cornerstone attribute of program and process and affects the cornerstone objective. Inadequate training of hazardous material workers regarding the reassembly and loading of shipping casks has a potential impact on public dose and on the licensees ability to safely package and transport radioactive material on public roadways. The violation involved an occurrence in the licensees radioactive material transportation program that is contrary to NRC or Department of Transportation regulations. When processed through the Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it: (1) was associated with radioactive material control, (2) involved the licensees program for radioactive material packaging and transportation, (3) did not cause radiation limits to be exceeded, (4) did not result in a breach of package during transit, (5) did not involve a certificate of compliance issue, (6) did not involve a non-compliance with low level burial ground, and (7) did not involve a failure to make notifications or to provide emergency information. In addition, this finding had cross-cutting aspects in the area of human performance in the component of resources because the licensee did not ensure the availability and adequacy of training for hazardous material workers involved in the shipment of radioactive material was available. Hazardous material workers are required to be trained and qualified to prepare radioactive material shipments for transport. (H.2.b) (Section 2PS2) Inspection Report# : 2007007 (pdf)

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: N/A Dec 30, 2007 Identified By: NRC Item Type: FIN Finding Identification and resolution of problems The inspectors reviewed approximately 200 condition reports, work orders, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities. The team concluded that the licensee was generally effective in identifying, evaluating, and correcting problems. Corrective actions, when specified, were generally implemented in a timely manner, although the team identified a significant number of longstanding equipment problems that were not being resolved in a timely manner. The team concluded that the licensee continued to have problems with the quality of operability assessments, and this was not being effectively addressed. The licensee performed quality higher-tier self-assessments, but the overall effectiveness was reduced by being slow to implement recommended improvements. The team concluded that the licensee was making progress in their efforts to address a trend in human performance, but this has not yet been completely effective. On the basis of 32 interviews conducted during this inspection, workers at the site felt free to report problems to their management, and were willing to use the corrective action program. An increased awareness and confidence in the Employee Concerns Program was also apparent. The team concluded that a positive safety-conscious work environment exists at Grand Gulf Nuclear Station. Inspection Report# : 2007008 (pdf) Last modified : August 29, 2008

Grand Gulf 1 3Q/2008 Plant Inspection Findings Initiating Events Significance: Jun 21, 2008 Identified By: Self-Revealing Item Type: FIN Finding Ineffective Corrective Actions in Response to Plant Transients Resulting from Animal Intrusions. The inspectors reviewed a self-revealing Green finding involving ineffective corrective actions that resulted in an unplanned down power caused by an animal intrusion. The plant experienced a loss of the balance of plant Transformer 23 with a loss of power to the plant service water pumps. Operators reduced reactor power to 47 percent. The control room dispatched operators to the river via a boat due to high river level and discovered a dead raccoon in the vicinity of the transformer. The inspectors noted that two previous reactor scrams had been caused by raccoons, and an injured raccoon had previously been found at the base of Transformer 23. The inspectors concluded that the flooding conditions which have been routinely experienced at the site and the occurrence of raccoon events at the site could have been used to anticipate and mitigate the unplanned down power. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2008-02089. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the Initiating Event and Mitigating Systems Cornerstone. The inspectors performed a Phase 2 analysis using Appendix A Determining the Significance of Reactor Inspection Findings for At-Power Situations, of Manual Chapter 0609, Significance Determination Process, and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors determined there was an increase in likelihood of a transient without the power conversion system but there was no reduction in remaining capability. Because the exposure time of the finding was less than 30 days, the result of the Phase 2 analysis was that the finding had very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with operating experience in that the licensee failed to implement proper corrective actions to prevent animals from causing a plant transient [P.2(b)]. Inspection Report# : 2008003 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: FIN Finding Ineffective Corrective Actions in Response to Resin in the Electro-hydraulic Control System. The inspectors identified a finding involving ineffective corrective actions in response to resin intrusion in the electro-hydraulic control system. The inspectors reviewed the corrective actions from a condition report involving a resin intrusion into the electro-hydraulic control system via a failed temporary ion-exchange filter in 2003. Review of the corrective actions associated with the 2003 event revealed that a long-range recovery plan was developed to remove resin from the electro-hydraulic control system. However, the recovery plan corrective actions were closed without licensee actions to remove resin from the electro-hydraulic control system. The failure to implement effective corrective actions following the 2003 resin intrusion event directly resulted in electro-hydraulic control stability issues seen in the fall of 2007, including reactor pressure perturbations and reductions in reactor power. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2007-04972. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 worksheet, the finding was determined to have very low safety significance because the finding did not contribute to the likelihood that mitigating equipment would not be available following a reactor trip. (Section 4OA2) Inspection Report# : 2008002 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Improper Control of Troubleshooting Causes a Loss of Condenser Vacuum The inspectors identified a finding involving a loss of condenser vacuum caused by improper troubleshooting of the seal steam pressure controller. Specifically, the licensee failed to provide adequate work instructions and procedural limitations during troubleshooting of the seal steam pressure controller. As a result, the plant experienced a loss of condenser vacuum and a plant transient. The licensee entered this issue

into their corrective action program as Condition Report CR-GGN-2007-04626. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the initiating event and mitigating systems cornerstone. The inspectors performed a Phase 2 analysis using Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," of Manual Chapter 0609, "Significance Determination Process," and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors assumed that only the power conversion system was affected and all other mitigating systems were available. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding was related to the human performance crosscutting component of decision making, in that the licensee failed to use conservative assumptions during troubleshooting activities and performed these activities without determining the validity of the troubleshooting instructions and identifying possible unintended consequences [H.1(b)] (Section 4OA3). Inspection Report# : 2007005 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Engineering Review of Plant Service Water Modification A self-revealing finding was identified involving the failure of a plant service water piping flange due to an improper flow control valve design modification. Specifically, the licensee failed to perform an adequate review of an engineering modification and the maintenance work orders did not have detailed installation instructions. As a result, the plant experienced a plant transient. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-05040. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have very low safety significance because the finding did not contribute to the likelihood that mitigating equipment or functions would not be available following a reactor trip. The cause of the finding was related to the human performance crosscutting component of work practices in that the responsible engineers failed to perform adequate self and peer checking during the development and review of the design modification to the plant service water flow control check valves [H.4(a)] (Section 4OA3). Inspection Report# : 2007005 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Failure to Follow Procedure Results in Loss of Condener Vacuum A self-revealing finding was identified involving a loss of condenser vacuum caused by plant operators improperly removing a steam jet air ejector from service. Specifically, the licensee failed to isolate the steam jet air ejector from service as delineated in the system operating instructions. As a result, the plant experienced a loss of condenser vacuum and a plant transient. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-05676. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the initiating event and mitigating systems cornerstone. The inspectors performed a Phase 2 analysis using Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," of Manual Chapter 0609, "Significance Determination Process," and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors assumed that only the power conversion system was affected and all other mitigating systems were available. Based on the results of the Phase 2 analysis, the finding was determined to have very low safety significance. The cause of the finding was related to the human performance crosscutting component of work practices in that the control room supervisor failed to ensure supervisory and management oversight of work activities such that nuclear safety is supported [H.4(c)] (Section 4OA2). Inspection Report# : 2007005 (pdf) Mitigating Systems Significance: Sep 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of Four Maintenance Rule Systems. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of four Maintenance Rule systems. Several discrepancies in the Maintenance Rule Program were discovered by the inspectors, including

unevaluated condition monitoring failures in the neutron monitoring system and an unevaluated functional failure in the standby gas treatment system. Plant personnel implemented additional corrective actions to fully investigate the potential extent of this condition and the apparent weakness in the condition report screening process used for the Maintenance Rule program. As a result, the Maintenance Rule expert panel classified four systems as needing increased monitoring and goal setting, moving these systems from an a(2) to an a(1) status. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-02219. This finding is more than minor since it was similar to Inspection Manual Chapter 0612, Appendix E, Example 7.b in that the problem involved degraded equipment performance. This finding was characterized under the significance determination process as having very low safety significance because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel failed to use proper self-checking and peer-checking to identify repetitive maintenance rule functional failures and also failed to properly document condition report screening activities [H.4(a)] (Section 1R12). Inspection Report# : 2008004 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment during Adverse Weather Conditions. The inspectors identified a Green noncited violation of 10 CFR 50.65 (a)(4), involving the failure to perform risk assessments following multiple declared tornado watches affecting Grand Gulf Nuclear Station during the landfall of Hurricane Gustav. On the morning of September 3, 2008, the inspectors noted that the licensee had not evaluated the increased risk from a declared tornado watch for the Claiborne County area. The inspectors brought this to the attention of plant personnel and a risk assessment was performed and plant risk was changed from a Green to a Yellow risk condition. The inspectors then reviewed the tornado watches declared by the National Weather Service that affected Claiborne County during the landfall of Hurricane Gustav, and noted that six separate tornado watches had been declared over the previous three days. A review of the control room logs showed no documentation of changes in plant risk condition. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-04397. This finding is more than minor because the risk assessments failed to consider unusual external conditions that were present or imminent. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, Assessment of Risk Deficit and consulting with the regional senior risk analyst, the inspectors determined the finding of very low safety significance due to a calculated incremental core damage probability deficit of 4.38E-08. This finding has a crosscutting aspect in the area of human performance associated with work practices in that plant personnel failed to follow the risk management procedure [H.4(b)] (Section 1R13). Inspection Report# : 2008004 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: VIO Violation Failure to Perform an Adequate Inspection of PMP Door Seals Protecting Safety Related Equipment. The inspectors identified a Green cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. The licensee had previously received a noncited violation for inadequate inspections of probable maximum precipitation door seals in NRC Inspection Report 05000416/2008002. On July 9, 2008, the inspectors found the entrance door to the Train B standby service water pump house not meeting the standards of the maintenance procedure because the door seals failed to make contact with the door. The extent of condition review found seven additional door seals degraded, including the doors to the diesel generator building and control building. The door seal on the Train B standby service water pump house identified by the inspectors on July 9, 2008, had not been identified by plant personnel during an extent of condition review on February 29, 2008. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-03216. The finding is more than minor since it affects the protection against external factors attribute of mitigating system cornerstone. The door seals also represent a degrading condition that if left uncorrected could affect the availability, reliability, and capability of mitigating systems required to respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multi-trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis using many bounding and conservative assumptions. The result was a delta-CDF of 3.3E 7/yr and a delta-LERF of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to take adequate corrective actions to ensure degraded probable maximum precipitation door seals were properly evaluated and repaired in a timely manner [P.1(d)] (Section 4OA2). Inspection Report# : 2008004 (pdf) Significance: Jun 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Fireproofing on Fire Barrier Protecting the Safeguards Switchgear Room

The inspectors identified a Green noncited violation of Facility Operating License Condition 2.C(41) involving the failure to ensure that fire barriers protecting safety-related areas were functional. The inspectors identified an 8-foot length of structural steel in the east stairwell wall, which is shared by the Division I safeguards switchgear room, that did not have the required fireproofing to maintain an adequate fire barrier. The missing passive fire protection reduced the fire rating of the wall by allowing heat to transfer through the unprotected steel, thus degrading the fire containment capability assumed in the fire hazards analysis. The licensee entered this issue into their corrective action program as Condition eport CR GGN 2008 01849. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire confinement category. The inspectors assigned a high degradation rating due to the fact that the required fireproofing was missing. The inspectors used the supplemental screening process for fire confinement findings and concluded that the finding was of very low safety significance (Green) due to the fact that the degraded barrier would have provided a minimum of 20 minutes fire endurance protection and there were no fire ignition sources or combustible materials in the area that would subject the barrier to direct flame impingement. Inspection Report# : 2008003 (pdf) Significance: Jun 21, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Recognize the Division III Diesel Generator being Non-Functional. The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1(a) involving the failure to follow a system operating instruction. While shutting down the Division III diesel generator, operators failed to place the outside air fan in automatic alignment resulting in the Division III diesel generator being nonfunctional. On May 5, 2008, operators had shutdown the Division III diesel generator, but they failed to recognize that the outside air fan was not running when they depressed the shutdown pushbutton for the outside air fan per the system operating instruction. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-02265 The finding is more than minor since it affects configuration control attribute of the Mitigating System Cornerstone objective, in that it affected the availability, reliability and capability of an onsite power source that supplies a bus that provides power to mitigating systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not represent a loss of a safety function that exceeded the Technical Specification allowed outage time. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices in that the operating crew did not use the proper human performance techniques of self checking while securing the outside air fan for the Division III diesel generator [H.4(a)]. Inspection Report# : 2008003 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate Fire Wrap Testing Discrepancies A noncited violation of License Condition 2.C(41), "Fire Protection Program," was identified because the licensee failed to evaluate vendor fire test results to ensure that deviations from the acceptance criteria did not result in a reduction in the effectiveness of the approved Fire Protection Program. The licensee replaced existing fire barrier material installed on conduits with 3M Interam fire wrap without recognizing that applicable NRC test criteria were not met. As a result, the licensee failed to perform an evaluation to determine whether the test results would result in a reduction in the effectiveness of the fire protection provided to the cables inside the affected conduits. The new fire wrap was installed to protect redundant trains of cables necessary for safe shutdown between 2004 and 2007. This finding was entered into the licensees corrective action program under Condition Report 2008-01910. The licensee took prompt compensatory measures and implemented hourly fire watches while the issue was being evaluated. Failure to properly evaluate vendor fire test results that did not satisfy the acceptance criteria in Generic Letter 86-10, Supplement 1 prior to changing the existing fire wrap with 3M Interam fire wrap as required by the approved Fire Protection Program was a performance deficiency. This finding was more than minor because it affected the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone Objective to ensure the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. This performance deficiency was also similar to the "more than minor" portion of Inspection Manual Chapter 0612, Appendix B, Example 3.i, in that an engineering evaluation was necessary to determine the acceptability of the existing fire wrap to perform its intended function. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected fire protection defense-in-depth strategies involving post-fire safe shutdown systems. This finding screened as having very low safety significance because it involved a fire barrier with a low degradation, since the nonconforming condition was subsequently determined to provide an acceptable margin to damage for the cables being protected. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC

Item Type: NCV NonCited Violation Fire Brigade Members Assigned Responsibilities That Conflicted with Fire Brigade Responsibilities. A noncited violation of License Condition 2.C.(41) was identified for failure to maintain required staffing available to respond to a fire. Specifically, the approved Fire Protection Program requires that a five-person fire brigade be available onsite at all times and not assigned duties that conflict with the duties of the fire brigade. Contrary to this, on three occasions in March 2008, operators assigned as fire brigade members were directed to perform operator rounds at the radial wells. Because the Mississippi River was at flood stage, this required traveling by boat, so the operators were unable to return to the plant promptly for approximately 2 hours. This was further complicated by the fact that operator/fire brigade radios did not work during most of the boat trip and in the vicinity of the most distant well, meaning that operators could not be quickly recalled. This finding was entered into the corrective action program under Condition Report 2008-01616. This finding had a crosscutting aspect in the area of Human Performance - Work Control (H.3.b) because the licensee did not ensure that different job activities were coordinated to ensure that the fire brigade remained available at all times. Failure to maintain a fully staffed fire brigade available onsite at all times was a performance deficiency. This finding was more than minor because it affected the protection from external factors (fire) attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesireable consequences. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected a fire protection defense-in-depth element. This finding was assigned a low degradation rating because the operations shift during the times when the fire brigade member was unavailable included extra fire brigade-trained personnel that could supplement the fire brigade. The delay in a replacement person reporting to the scene of a fire would not have impacted the initial fire fighting effort, since enough fire brigade personnel were available to perform the functions. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Plant Modification Blocked Access for Manual Firefighting. A noncited violation of License Condition 2.C.(41), "Fire Protection Program," was identified related to making a plant change that negatively impacted the effectiveness of the approved Fire Protection Program. The team identified that the licensee had permanently blocked the door to the abandoned Unit 2 portion of the joint control room without performing a fire protection impact evaluation. The only remaining access path was a small hatch that would have made it difficult for fire fighters to gain access with protective clothing and equipment. A fire in this area could threaten operations in the Unit 1 control room if not promptly suppressed. This finding was determined to have a cross-cutting aspect in problem identification and resolution timeliness (P.1.d) because fire protection personnel recognized that a new access door was needed in 2006, but no substantial action had been taken to install it by the time of this inspection. This finding was entered in to the licensees corrective action program under Condition Reports 2008-001893 and 2008-01913. Blocking access to the Unit 2 control room area and not promptly restoring access to allow manual fire suppression was a performance deficiency. This finding was more than minor because it affected the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected a fire protection defense-in-depth element. This finding was determined to have very low safety significance because all potential fire ignition sources in the affected area screened out in Task 2.3.4 in the Phase 2 evaluation. There were no ignition sources because the licensee had removed electrical power from this area, and administratively prevented hot work and storage of transient combustible material from this area. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure That Potential Damage to Motor-Operated Valve Circuits Would Not Prevent Safe Shutdown. A noncited violation of 10 CFR Part 50, Appendix R, Section III.G.1.a was identified because the licensee failed to evaluate the impact of a potential motor operated valve failure mechanism on the ability to implement post-fire safe shutdown following a control room evacuation. The team identified that the Residual Heat Removal Pump Minimum Flow Valve F064A could be damaged by fire in the control room and not be available to perform its safe shutdown function. This finding involved mechanistic damage due to hot shorts as described in Information Notice 92-18, "Potential for Loss of Remote Shutdown Capability During Control Room Fire." The licensee had incorrectly interpreted this operating experience and concluded that no action was required. This finding was entered into the corrective action program under Condition Reports 1999-0236 and 2008-01904. The team determined that failure to ensure that components necessary to safely shutdown the reactor would remain operable following a fire was a performance deficiency. This deficiency was more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to external events (fire) to prevent undesirable consequences. The Phase 3 risk evaluation performed by the senior reactor analyst determined this deficiency had very low safety significance because the probability of having a fire in either of the two control room panels where the postulated damage could occur and lead to a control room evacuation was very low. Inspection Report# : 2008006 (pdf)

Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation (PMP) Door Seals Protecting Safety Related Equipment. The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. The inspectors identified that the door seals did not make contact with the door frame and the door had a significant amount of corrosion underneath the door seals, indicating long term degradation. The extent of condition review found three additional door seals with degraded conditions, including doors to the standby service water pump houses. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2008-01123 and 2008-01623. This finding was more than minor because the door seals represent a degrading condition that if left uncorrected could become a more significant safety concern. The inspectors determined this finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance since it did not represent an actual loss of safety function for the standby service water pumps or the diesel generators. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to properly identify the degraded conditions of the probable maximum precipitation door seals during their surveillance inspection. [P.1(a)] (Section 1R01) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement an Adequate Compensatory Fire Watch per Station Fire Protection Procedures. The inspectors identified a noncited violation of Facility Operating License Condition 2.C.41 for the failure to properly implement a compensatory fire watch per the station fire protection program. The inspectors performed a fire inspection of the auxiliary building electrical penetration room. The inspectors noted that plant personnel had not entered the room to perform a required fire watch. The inspectors questioned security personnel, reviewed the fire watch log and determined that the fire watch log had been checked off as completed. The completion time corresponded to the time the inspector was in the room. After further review and interviews with security personnel, the inspectors determined that the plant employee designated to perform the fire watch duties misunderstood the requirements for the fire watch. The employee had only verified the auxiliary building hallway area outside the room and failed to check inside the auxiliary building electrical penetration room as required. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-00869. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for outages or degradations. The inspectors assigned a high degradation rating due to the fact that automatic fire suppression system was tagged out and inoperable. Because the system was degraded without compensatory actions for approximately 2 hours, the inspectors used a duration factor of 0.01. The inspectors used 2E-2 for a generic fire frequency area which corresponds to Table 1.4.2, Generic Fire Area Fire Frequencies for a switchgear room. The resulting calculated change in core damage frequency was 2E-4, which was greater than the high degradation Phase 1 Quantitative Screening Criteria of 1E-6, requiring a Phase 2 analysis. The inspectors consulted with a regional Senior Reactor Analyst and a simplified Phase 3 was performed using a duration factor of 2.3E-4 for the 2-hour time period, and the IPEEE specific room fire frequency of 7.2E-4. The resulting calculated change in core damage frequency was 1.7E-7, which would be less that the Phase 1 quantitative screening criteria. Using this information, the regional Senior Reactor Analyst, determined the finding to be of very low safety significance. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices in that the individual assigned to perform the fire watch proceeded in the face of uncertainty and failed to use appropriate human error prevention techniques. [H.4(a)] (Section 1R05) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure of Licensed Senior Reactor Operators to Maintain the Required Proficiency to Maintain Their License Current. The inspectors identified a noncited violation of 10 CFR 55.53.e, Conditions of License, for failure of licensed senior reactor operators to maintain the required proficiency to maintain their license current. Senior reactor operators standing the shift supervisor/shift technical advisor position were taking credit for senior reactor operator proficiency watches while standing this position. The normal shift complement of senior reactor operators consist of a shift manager, a control room supervisor, and a shift supervisor/shift technical advisor. When this issue was brought to the attention of operations management; they stopped the practice of the shift supervisor/shift technical advisor receiving senior reactor operator proficiency watch credit for standing that position. All shift supervisor/shift technical advisor senior reactor operators were inactivated. The plant issued a standing order that prohibited the shift supervisor/shift technical advisor to be allowed to perform the

senior reactor operators oversight function in the control room and the shift manager or control room supervisor had to be in the control room at all times. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-01126. This finding was more than minor because if left uncorrected the finding could become a more significant safety concern. This finding affects the mitigating system cornerstone. The finding was determined to be of very low safety significance using the Licensed Operator Requalification Significance Determination Process since it related to operator license conditions and more than 20 percent of the affected individuals were deficient (Section 1R11). Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform a Required Technical Specification Surveillance. The inspectors identified a noncited violation of Technical Specifications 3.8.1, AC Sources-Operating, for the failure to perform a required surveillance following the loss of a required offsite power source. The plant suffered a loss of power from the Port Gibson 115 kV line during high winds. Due to the fact that there is no direct control room alarm to alert the operating crew, they were not immediately aware they had lost the offsite source of power. When the crew recognized the loss of the bus they only entered a potential limiting condition of operations, due to the crew failing to realize that this was one of the required offsite sources. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2008-00737 and 2008-01202. This finding was more than minor because it impacts the mitigating system cornerstone objective in that it affects the operability, availability, reliability of an offsite power source that supplies a bus that provides power to mitigating systems. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was of very low safety significance since it did not represent an actual loss of a safety function. The cause of this finding has a crosscutting aspect in the area of human performance associated with the resources attribute in that the operators did not have adequate procedural guidance to determine the loss of safety-related offsite power supply. [H.2(c)] (Section 1R22) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate Cracks in Standby Service Water Pump House Structure. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, for failing to implement effective corrective actions after identifying concrete cracking in the standby service water pump houses. The inspectors determined that the program that evaluates, monitors, and repairs cracks for all safety related structures only identified a single crack for the entire site and does not track other structural cracks previously identified in the corrective action program. The last program inspection had been performed as recently as October 25, 2007, and only identified the single crack that had been documented in previous inspections. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2007-05824. This finding was more than minor because the cracks represent a degrading condition that if left uncorrected could become more significant safety concern. The inspectors determined this finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was of very low safety significance since it did not represent an actual loss of a safety function. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee personnel failed to properly maintain and utilize the program for evaluating, tracking and repairing identified concrete cracks in safety related structures. [H.4(b)] (Section 4OA2) Inspection Report# : 2008002 (pdf) Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Venting Procedure for the Reactor Core Isolation Cooling System The inspectors identified a noncited violation of Criterion V, Instructions, Procedures, and Drawings, of 10 CFR Part 50, Appendix B for the failure to demonstrate compliance with Technical Specification Surveillance Requirement 3.5.3.1 due to an inadequate surveillance procedure. The reactor core isolation cooling system is vented at the injection valve through a hard-piped drain with no visual means of detecting air in the system. The inspectors determined that the procedure failed to contain adequate acceptance criteria to qualitatively or quantitatively assess abnormal amounts of air in the reactor core isolation cooling system. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2007-03818. The finding was greater than minor because it affects the procedure quality 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. Using the Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, the finding was determined to have a very low safety significance in that it did not result in the actual loss of the reactor core isolation cooling system, and was not potentially risk-significant due to external initiating events. Inspection Report# : 2007005 (pdf)

Significance: Dec 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform a root cause analysis for RHR heat exchanger B fouling, and implement corrective action to prevent recurrence A noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, was identified for failure to perform an adequate cause analysis for fouling of the Residual Heat Removal Heat Exchanger B on the standby service water side, and implement corrective action to prevent recurrence. This fouling reduced the thermal performance margin to 0.6 percent, but was not treated as a significant condition adverse to quality within the corrective action program. The licensee chose to temporarily restore margin by increasing the flow rate, but this did not remove or stop the fouling from continuing to occur. This finding has cross cutting aspects in the decision-making area of Human Performance (H.1.b) because the licensees decision-making in response to this degraded condition did not use conservative criteria in deciding when to clean this heat exchanger, and did not verify that the underlying assumptions remained valid. Failure to treat Residual Heat Removal Heat Exchanger B degradation as a significant condition adverse to quality, and perform an adequate cause analysis, and implement corrective action to prevent recurrence was a performance deficiency. This was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the system could become fouled enough to prevent removing the required heat load without the licensee recognizing this condition. This finding affected the Mitigating Systems and Barrier Integrity Cornerstones, since this component was required for both decay heat removal and containment heat removal functions. In accordance with the Phase 1 Significance Determination Process instructions, the significance was assessed using the Mitigating Systems Cornerstone, since this represented the dominant risk. This finding was determined to have very low safety significance (Green) during a Phase 1 Significance Determination Process, since it was confirmed to not involve loss of the design heat removal capability. This issue was entered into the licensees corrective action program under Condition Report 2007-5766. (Section 4OA2.e.1(b)(1)) Inspection Report# : 2007008 (pdf) Significance: Dec 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate thermal performance testing of the residual heat removal heat exchangers A noncited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, was identified because the licensees thermal performance test procedures for the residual heat removal heat exchangers were inadequate to ensure the quality of the test results. Specifically, the test procedure failed to specify adequate prerequisites for minimum heat load and use of high-accuracy instrumentation. This resulted in test results used to meet commitments for the Generic Letter 89-13 test program which provided little useful information due to high inaccuracy. Failure to adequately test and trend the thermal performance of the residual heat removal heat exchangers was a performance deficiency because it masked the actual thermal performance to the point where the licensee did not recognize the onset of fouling. The team determined that these heat exchangers began to experience fouling between 1997 and 1998, but this was not recognized. In the case of Residual Heat Removal Heat Exchanger B, the degraded performance was determined to be sufficient to make the fouling factor exceed the design value, necessitating compensatory action to be able to show continued operability. This was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the system could become fouled enough to prevent removing the required heat load without the licensee recognizing this condition. This finding affected the Mitigating Systems and Barrier Integrity Cornerstones, since this component was required for both decay heat removal and containment heat removal functions. In accordance with the Phase 1 SDP instructions, the significance was assessed using the Mitigating Systems Cornerstone, since this represented the dominant risk. This finding was determined to have very low safety significance (Green) during a Phase 1 Significance Determination Process, since it was confirmed to not involve loss of the design heat removal capability. This issue was entered into the licensees corrective action program under Condition Report 2008-0412. (Section 1R07) Inspection Report# : 2007008 (pdf) Barrier Integrity Significance: Jun 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Causing a Loss of Decay Heat Removal to the Spent Fuel Pool. The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure of operators to follow a safety-related off normal event procedure resulting in a loss of decay heat removal to the spent fuel pool. The operators elected to remove cooling to the fuel pool cooling heat exchangers to minimize the temperature rise on the component cooling water system during a partial loss of the plant service water system. This action was not specified in the off-normal event procedure. The off-normal event procedure only permitted the isolation of component cooling water flow to the fuel pool cooling heat exchangers for degraded component cooling water flow or pressure. This resulted in the spent fuel pool losing decay heat removal for approximately 3 hours and 22 minutes. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2008-02147. The finding is more than minor since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone

objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not preclude operators from restoring spent fuel pool cooling to ensure the Fuel Barrier Cornerstone. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making in that operators did not use a systematic decision making process when faced with unexpected plant conditions [H.1(a)]. Inspection Report# : 2008003 (pdf) Significance: Mar 22, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Design Control of HPCS Minimum Flow Valve Motor-Operated Valve Over Current Setpoint. The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to properly set the over current trip setpoint for the high pressure core spray minimum flow motor operated valve. This resulted in a spurious over current trip of the valve breaker during a high pressure core spray momentary pump start for breaker operability following post Division 3 emergency core cooling system testing. As a result of the trip, the high pressure core spray minimum flow valve failed open. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-01201. The finding was more than minor because it was associated with the barrier integrity cornerstone to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the MC 0609, "Significance Determination Process," Phase 1 worksheet, the finding was determined to have very low safety significance since it did not result in a loss of the containment barrier. Additionally, the issue was screened and determined to not impact the High Pressure Core Spray mitigating system function. (Section 4OA3) Inspection Report# : 2008002 (pdf) Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2007 Identified By: NRC Item Type: FIN Finding Inadequate Procedure The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for the failure to provide a detailed work order package to perform vent and fill operations on a pressure transmitter. Specifically, the licensee did not provide appropriate instructions in a work order package to properly isolate pressure Transmitter 1N64N006B prior to opening the drain valve. Consequently, this resulted in the release of radioactive gas from the system and an unplanned and unintended exposure for two individuals involved in the work activity. The finding is more than minor because it is associated with the occupational radiation safety attribute of program and process and affected the cornerstone objective because it involved unplanned and unintended dose to two workers. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance because: it did not involve: (1) as low as reasonably achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. In addition, this finding has a cross-cutting aspect in the area of work control associated with work planning because the licensee failed to properly plan work activities by incorporating specific plant system details into the work order to allow the instrumentation and control technicians to properly drain a pressure transmitter [H.3(a)] (Section 2OS2). Inspection Report# : 2007005 (pdf) Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security

cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: N/A Dec 30, 2007 Identified By: NRC Item Type: FIN Finding Identification and resolution of problems The inspectors reviewed approximately 200 condition reports, work orders, associated root and apparent cause evaluations, and other supporting documentation to assess problem identification and resolution activities. The team concluded that the licensee was generally effective in identifying, evaluating, and correcting problems. Corrective actions, when specified, were generally implemented in a timely manner, although the team identified a significant number of longstanding equipment problems that were not being resolved in a timely manner. The team concluded that the licensee continued to have problems with the quality of operability assessments, and this was not being effectively addressed. The licensee performed quality higher-tier self-assessments, but the overall effectiveness was reduced by being slow to implement recommended improvements. The team concluded that the licensee was making progress in their efforts to address a trend in human performance, but this has not yet been completely effective. On the basis of 32 interviews conducted during this inspection, workers at the site felt free to report problems to their management, and were willing to use the corrective action program. An increased awareness and confidence in the Employee Concerns Program was also apparent. The team concluded that a positive safety-conscious work environment exists at Grand Gulf Nuclear Station. Inspection Report# : 2007008 (pdf) Last modified : November 26, 2008

Grand Gulf 1 4Q/2008 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Trip of a Reactor Recirculation Pump During Pump Up-shift to Fast Speed Due to Ineffective Corrective Actions The inspectors reviewed a self-revealing Green finding involving a recirculation pump trip during pump up-shift to fast speed due to ineffective corrective actions. The plant had recently replaced the recirculation motor on Pump A during the refuelling outage and during investigation determined that the instantaneous over-current trip for the breaker had drifted low. The inspectors performed a review of condition reports and determined that reactor recirculation Pump B had tripped following motor replacement for the same reason in September 2007. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-06269. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not contribute to loss of function of mitigating equipment. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program in that the licensee failed to perform a thorough evaluation of a problem that resulted in a plant transient such that the resolution properly addressed the cause and extent of condition [P.1(c)]. (Section 1R20) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Automatic Reactor Scram Caused by an Operator Inadvertently Closing Steam Supply Valves to the Reactor Feed Pump Turbine The inspectors reviewed a self-revealing Green finding involving an automatic reactor scram caused by an operator inadvertently closing steam supply valves to the reactor feed pump turbine. Site personnel investigating the scram determined that an operator had incorrectly performed actions for the reactor feed Pump B turbine on the reactor feed Pump A turbine control switches at a local panel. The operator inadvertently closed the steam supply valves to the reactor feed Pump A turbine resulting in a total loss of feedwater flow and low reactor water level scram. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-06195. The finding was more than minor because it was associated with the initiating events cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that an evaluation was required by the regional senior reactor analyst, because the finding impacted both the initiating event and mitigating systems cornerstone. The senior reactor analyst performed a Phase 3 analysis and determined the issue was very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the operator failed to use proper self-checking techniques while performing actions to place feed Pump B in the standby lineup [H.4(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf)

Significance: Jun 21, 2008 Identified By: Self-Revealing Item Type: FIN Finding Ineffective Corrective Actions in Response to Plant Transients Resulting from Animal Intrusions. The inspectors reviewed a self-revealing Green finding involving ineffective corrective actions that resulted in an unplanned down power caused by an animal intrusion. The plant experienced a loss of the balance of plant Transformer 23 with a loss of power to the plant service water pumps. Operators reduced reactor power to 47 percent. The control room dispatched operators to the river via a boat due to high river level and discovered a dead raccoon in the vicinity of the transformer. The inspectors noted that two previous reactor scrams had been caused by raccoons, and an injured raccoon had previously been found at the base of Transformer 23. The inspectors concluded that the flooding conditions which have been routinely experienced at the site and the occurrence of raccoon events at the site could have been used to anticipate and mitigate the unplanned down power. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2008-02089. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the Initiating Event and Mitigating Systems Cornerstone. The inspectors performed a Phase 2 analysis using Appendix A Determining the Significance of Reactor Inspection Findings for At-Power Situations, of Manual Chapter 0609, Significance Determination Process, and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors determined there was an increase in likelihood of a transient without the power conversion system but there was no reduction in remaining capability. Because the exposure time of the finding was less than 30 days, the result of the Phase 2 analysis was that the finding had very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with operating experience in that the licensee failed to implement proper corrective actions to prevent animals from causing a plant transient [P.2(b)]. Inspection Report# : 2008003 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: FIN Finding Ineffective Corrective Actions in Response to Resin in the Electro-hydraulic Control System. The inspectors identified a finding involving ineffective corrective actions in response to resin intrusion in the electro-hydraulic control system. The inspectors reviewed the corrective actions from a condition report involving a resin intrusion into the electro-hydraulic control system via a failed temporary ion-exchange filter in 2003. Review of the corrective actions associated with the 2003 event revealed that a long-range recovery plan was developed to remove resin from the electro-hydraulic control system. However, the recovery plan corrective actions were closed without licensee actions to remove resin from the electro-hydraulic control system. The failure to implement effective corrective actions following the 2003 resin intrusion event directly resulted in electro-hydraulic control stability issues seen in the fall of 2007, including reactor pressure perturbations and reductions in reactor power. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2007-04972. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the MC 0609, "Significance Determination Process," Phase 1 worksheet, the finding was determined to have very low safety significance because the finding did not contribute to the likelihood that mitigating equipment would not be available following a reactor trip. (Section 4OA2) Inspection Report# : 2008002 (pdf)

Mitigating Systems Significance: Dec 31, 2008 Identified By: NRC Item Type: FIN Finding Inadequate Fire Drill Critique The inspectors identified a finding for fire brigade performance deficiencies that were not identified by the licensee during a fire drill critique. The inspectors identified several deficiencies during the drill including issues relating to command and control, fire fighting strategy and use of fire fighting equipment. The inspectors provided feedback to plant personnel on the identified performance issues and the inadequate drill evaluation. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 06522. This finding was more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone objective and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, was used to analyze the finding since the inadequate critique had an adverse effect on fire brigade effectiveness, in relation to defense-in-depth strategies. Manual Chapter 0609, Appendix F states that findings associated with the onsite manual fire brigade are excluded. Therefore, in accordance with Manual Chapter 0609, the safety significance was determined by regional management review. Regional management concluded that the finding was of very low safety significance because it reflected fire brigade performance during a training drill, rather than during an actual fire. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to have a low enough threshold in identifying performance issues associated with a plant fire drill [P.1(a)]. (Section 1R05) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of the Engineering Safety Features Electrical Switchgear and Battery Room Ventilation System The inspectors identified a noncited violation of 10 CFR 50.65(a)(2) for the failure to adequately monitor the performance of the engineering safety features electrical switchgear and battery room ventilation system. The inspectors identified a condition report from March 2004 that had not been screened and evaluated in the maintenance rule database as a maintenance preventable functional failure. The condition report identified a room cooler that had tripped due to excessive current on the fan motor because an incorrectly sized sheave was installed during previous maintenance. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 02219. The inspectors determined that this finding was more than minor since the engineering safety features electrical switchgear and battery room ventilation system was not placed in (a)(1) monitoring status in a timely manner. In addition, the finding was more than minor since violations of 10 CFR 50.65(a)(2) necessarily involve degraded system performance, which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the maintenance rule aspect of the finding did not lead to an actual loss of safety function of the system or cause a component to be inoperable, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. (Section 1R12) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of Inadequate Operability Evaluations The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving two

examples of a failure to follow procedures which resulted in inadequate operability evaluations. The first example involved an inadequate evaluation of foreign material in the condensate storage tank. The evaluation relied on an assumption that the high-pressure core spray and reactor core isolation cooling pumps would not be damaged by metal debris entrained in the pumps suction. The second example involved an inadequate evaluation of the structural integrity of the standby service water cooling towers that only considered the loss of structural support from a single beam. The licensee entered these issues into the corrective action program as Condition Reports CR GGN 2008 05685 and CR GGN 2008 06044. This finding is more than minor because the failure to perform adequate operability evaluations, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions [H.1(b)]. (Section 1R15 Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Standby Service Water Corrosion The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving a failure to take corrective actions to prevent recurrence of severe corrosion in piping hangers, piping supports, and piping in the standby service water basin cooling towers. Significant corrosion of the standby service water supports in October 2008 had been previously identified by plant personnel during a ten-year in-service inspection on October 3, 1993. At that time, plant personnel determined this to be a significant degraded condition of a safety related system, requiring replacement of the piping and associated supports. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-05434. This finding was more than minor because the corrosion represented a degrading condition that if left uncorrected could become more significant safety concern. The finding was also more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not represent an actual loss of safety function, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Following Identification of Degrading Standby Service Water Supports The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, involving the failure to take timely corrective actions for corrosion on distribution beam structural support posts in the standby service water basin cooling towers. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-05434. The finding was more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance since it did not represent an actual loss of safety function of the standby service water cooling towers, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding has a crosscutting aspect in the area of problem

identification and resolution associated with the corrective action program because licensee personnel failed to identify issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Nov 06, 2008 Identified By: NRC Item Type: NCV NonCited Violation B.5.b. Phase 2 and 3 Mitigating Strategy This finding, affecting the Mitigating Systems Cornerstone, is related to mitigative measures developed to cope with losses of large areas of the plant; in response to Section B.5.b. of the February 25, 2002, Interim Compensatory Measures (ICM) Order (EA-02-026) and related NRC guidance. This finding has been designated as "Official Use Only - Security-Related Information;" therefore, the details of this finding are being withheld from public disclosure. This finding has no cross-cutting aspect. See inspection report 2008-007 for more details. Inspection Report# : 2008007 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of Four Maintenance Rule Systems. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of four Maintenance Rule systems. Several discrepancies in the Maintenance Rule Program were discovered by the inspectors, including unevaluated condition monitoring failures in the neutron monitoring system and an unevaluated functional failure in the standby gas treatment system. Plant personnel implemented additional corrective actions to fully investigate the potential extent of this condition and the apparent weakness in the condition report screening process used for the Maintenance Rule program. As a result, the Maintenance Rule expert panel classified four systems as needing increased monitoring and goal setting, moving these systems from an a(2) to an a(1) status. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-02219. This finding is more than minor since it was similar to Inspection Manual Chapter 0612, Appendix E, Example 7.b in that the problem involved degraded equipment performance. This finding was characterized under the significance determination process as having very low safety significance because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel failed to use proper self-checking and peer-checking to identify repetitive maintenance rule functional failures and also failed to properly document condition report screening activities [H.4(a)] (Section 1R12). Inspection Report# : 2008004 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment during Adverse Weather Conditions. The inspectors identified a Green noncited violation of 10 CFR 50.65 (a)(4), involving the failure to perform risk assessments following multiple declared tornado watches affecting Grand Gulf Nuclear Station during the landfall of Hurricane Gustav. On the morning of September 3, 2008, the inspectors noted that the licensee had not evaluated the increased risk from a declared tornado watch for the Claiborne County area. The inspectors brought this to the attention of plant personnel and a risk assessment was performed and plant risk was changed from a Green to a Yellow risk condition. The inspectors then reviewed the tornado watches declared by the National Weather Service that affected Claiborne County during the landfall of Hurricane Gustav, and noted that six separate tornado watches had been declared over the previous three days. A review of the control room logs showed no documentation of changes in plant risk condition. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-04397.

This finding is more than minor because the risk assessments failed to consider unusual external conditions that were present or imminent. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, Assessment of Risk Deficit and consulting with the regional senior risk analyst, the inspectors determined the finding of very low safety significance due to a calculated incremental core damage probability deficit of 4.38E-08. This finding has a crosscutting aspect in the area of human performance associated with work practices in that plant personnel failed to follow the risk management procedure [H.4(b)] (Section 1R13). Inspection Report# : 2008004 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: VIO Violation Failure to Perform an Adequate Inspection of PMP Door Seals Protecting Safety Related Equipment. The inspectors identified a Green cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. The licensee had previously received a noncited violation for inadequate inspections of probable maximum precipitation door seals in NRC Inspection Report 05000416/2008002. On July 9, 2008, the inspectors found the entrance door to the Train B standby service water pump house not meeting the standards of the maintenance procedure because the door seals failed to make contact with the door. The extent of condition review found seven additional door seals degraded, including the doors to the diesel generator building and control building. The door seal on the Train B standby service water pump house identified by the inspectors on July 9, 2008, had not been identified by plant personnel during an extent of condition review on February 29, 2008. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-03216. The finding is more than minor since it affects the protection against external factors attribute of mitigating system cornerstone. The door seals also represent a degrading condition that if left uncorrected could affect the availability, reliability, and capability of mitigating systems required to respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multi-trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis using many bounding and conservative assumptions. The result was a delta-CDF of 3.3E 7/yr and a delta-LERF of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to take adequate corrective actions to ensure degraded probable maximum precipitation door seals were properly evaluated and repaired in a timely manner [P.1(d)] (Section 4OA2). Inspection Report# : 2008004 (pdf) Significance: Jun 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Fireproofing on Fire Barrier Protecting the Safeguards Switchgear Room The inspectors identified a Green noncited violation of Facility Operating License Condition 2.C(41) involving the failure to ensure that fire barriers protecting safety-related areas were functional. The inspectors identified an 8-foot length of structural steel in the east stairwell wall, which is shared by the Division I safeguards switchgear room, that did not have the required fireproofing to maintain an adequate fire barrier. The missing passive fire protection reduced the fire rating of the wall by allowing heat to transfer through the unprotected steel, thus degrading the fire containment capability assumed in the fire hazards analysis. The licensee entered this issue into their corrective action program as Condition eport CR GGN 2008 01849. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the

inspectors determined that the finding impacted the fire confinement category. The inspectors assigned a high degradation rating due to the fact that the required fireproofing was missing. The inspectors used the supplemental screening process for fire confinement findings and concluded that the finding was of very low safety significance (Green) due to the fact that the degraded barrier would have provided a minimum of 20 minutes fire endurance protection and there were no fire ignition sources or combustible materials in the area that would subject the barrier to direct flame impingement. Inspection Report# : 2008003 (pdf) Significance: Jun 21, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Recognize the Division III Diesel Generator being Non-Functional. The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1(a) involving the failure to follow a system operating instruction. While shutting down the Division III diesel generator, operators failed to place the outside air fan in automatic alignment resulting in the Division III diesel generator being nonfunctional. On May 5, 2008, operators had shutdown the Division III diesel generator, but they failed to recognize that the outside air fan was not running when they depressed the shutdown pushbutton for the outside air fan per the system operating instruction. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-02265 The finding is more than minor since it affects configuration control attribute of the Mitigating System Cornerstone objective, in that it affected the availability, reliability and capability of an onsite power source that supplies a bus that provides power to mitigating systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not represent a loss of a safety function that exceeded the Technical Specification allowed outage time. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices in that the operating crew did not use the proper human performance techniques of self checking while securing the outside air fan for the Division III diesel generator [H.4(a)]. Inspection Report# : 2008003 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate Fire Wrap Testing Discrepancies A noncited violation of License Condition 2.C(41), "Fire Protection Program," was identified because the licensee failed to evaluate vendor fire test results to ensure that deviations from the acceptance criteria did not result in a reduction in the effectiveness of the approved Fire Protection Program. The licensee replaced existing fire barrier material installed on conduits with 3M Interam fire wrap without recognizing that applicable NRC test criteria were not met. As a result, the licensee failed to perform an evaluation to determine whether the test results would result in a reduction in the effectiveness of the fire protection provided to the cables inside the affected conduits. The new fire wrap was installed to protect redundant trains of cables necessary for safe shutdown between 2004 and 2007. This finding was entered into the licensees corrective action program under Condition Report 2008-01910. The licensee took prompt compensatory measures and implemented hourly fire watches while the issue was being evaluated. Failure to properly evaluate vendor fire test results that did not satisfy the acceptance criteria in Generic Letter 86-10, Supplement 1 prior to changing the existing fire wrap with 3M Interam fire wrap as required by the approved Fire Protection Program was a performance deficiency. This finding was more than minor because it affected the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone Objective to ensure the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. This performance deficiency was also similar to the "more than minor" portion of Inspection Manual Chapter 0612, Appendix B, Example 3.i, in that an engineering evaluation was necessary to determine the acceptability of the existing fire wrap to perform its intended function. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected fire protection defense-in-depth strategies involving post-fire safe shutdown systems. This finding screened as having very low safety significance because it involved a fire barrier with a low degradation, since the nonconforming condition

was subsequently determined to provide an acceptable margin to damage for the cables being protected. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Fire Brigade Members Assigned Responsibilities That Conflicted with Fire Brigade Responsibilities. A noncited violation of License Condition 2.C.(41) was identified for failure to maintain required staffing available to respond to a fire. Specifically, the approved Fire Protection Program requires that a five-person fire brigade be available onsite at all times and not assigned duties that conflict with the duties of the fire brigade. Contrary to this, on three occasions in March 2008, operators assigned as fire brigade members were directed to perform operator rounds at the radial wells. Because the Mississippi River was at flood stage, this required traveling by boat, so the operators were unable to return to the plant promptly for approximately 2 hours. This was further complicated by the fact that operator/fire brigade radios did not work during most of the boat trip and in the vicinity of the most distant well, meaning that operators could not be quickly recalled. This finding was entered into the corrective action program under Condition Report 2008-01616. This finding had a crosscutting aspect in the area of Human Performance - Work Control (H.3.b) because the licensee did not ensure that different job activities were coordinated to ensure that the fire brigade remained available at all times. Failure to maintain a fully staffed fire brigade available onsite at all times was a performance deficiency. This finding was more than minor because it affected the protection from external factors (fire) attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesireable consequences. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected a fire protection defense-in-depth element. This finding was assigned a low degradation rating because the operations shift during the times when the fire brigade member was unavailable included extra fire brigade-trained personnel that could supplement the fire brigade. The delay in a replacement person reporting to the scene of a fire would not have impacted the initial fire fighting effort, since enough fire brigade personnel were available to perform the functions. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Plant Modification Blocked Access for Manual Firefighting. A noncited violation of License Condition 2.C.(41), "Fire Protection Program," was identified related to making a plant change that negatively impacted the effectiveness of the approved Fire Protection Program. The team identified that the licensee had permanently blocked the door to the abandoned Unit 2 portion of the joint control room without performing a fire protection impact evaluation. The only remaining access path was a small hatch that would have made it difficult for fire fighters to gain access with protective clothing and equipment. A fire in this area could threaten operations in the Unit 1 control room if not promptly suppressed. This finding was determined to have a cross-cutting aspect in problem identification and resolution timeliness (P.1.d) because fire protection personnel recognized that a new access door was needed in 2006, but no substantial action had been taken to install it by the time of this inspection. This finding was entered in to the licensees corrective action program under Condition Reports 2008-001893 and 2008-01913. Blocking access to the Unit 2 control room area and not promptly restoring access to allow manual fire suppression was a performance deficiency. This finding was more than minor because it affected the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected a fire protection defense-in-depth element. This finding was determined to have very low safety significance because all potential fire ignition sources in the affected area screened out in Task 2.3.4 in the Phase 2 evaluation. There were no ignition sources because the licensee had removed electrical power from this area, and administratively prevented hot work and storage of transient combustible material from this area.

Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure That Potential Damage to Motor-Operated Valve Circuits Would Not Prevent Safe Shutdown. A noncited violation of 10 CFR Part 50, Appendix R, Section III.G.1.a was identified because the licensee failed to evaluate the impact of a potential motor operated valve failure mechanism on the ability to implement post-fire safe shutdown following a control room evacuation. The team identified that the Residual Heat Removal Pump Minimum Flow Valve F064A could be damaged by fire in the control room and not be available to perform its safe shutdown function. This finding involved mechanistic damage due to hot shorts as described in Information Notice 92-18, "Potential for Loss of Remote Shutdown Capability During Control Room Fire." The licensee had incorrectly interpreted this operating experience and concluded that no action was required. This finding was entered into the corrective action program under Condition Reports 1999-0236 and 2008-01904. The team determined that failure to ensure that components necessary to safely shutdown the reactor would remain operable following a fire was a performance deficiency. This deficiency was more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to external events (fire) to prevent undesirable consequences. The Phase 3 risk evaluation performed by the senior reactor analyst determined this deficiency had very low safety significance because the probability of having a fire in either of the two control room panels where the postulated damage could occur and lead to a control room evacuation was very low. Inspection Report# : 2008006 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation (PMP) Door Seals Protecting Safety Related Equipment. The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. The inspectors identified that the door seals did not make contact with the door frame and the door had a significant amount of corrosion underneath the door seals, indicating long term degradation. The extent of condition review found three additional door seals with degraded conditions, including doors to the standby service water pump houses. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2008-01123 and 2008-01623. This finding was more than minor because the door seals represent a degrading condition that if left uncorrected could become a more significant safety concern. The inspectors determined this finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was determined to have very low safety significance since it did not represent an actual loss of safety function for the standby service water pumps or the diesel generators. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to properly identify the degraded conditions of the probable maximum precipitation door seals during their surveillance inspection. [P.1(a)] (Section 1R01) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement an Adequate Compensatory Fire Watch per Station Fire Protection Procedures.

The inspectors identified a noncited violation of Facility Operating License Condition 2.C.41 for the failure to properly implement a compensatory fire watch per the station fire protection program. The inspectors performed a fire inspection of the auxiliary building electrical penetration room. The inspectors noted that plant personnel had not entered the room to perform a required fire watch. The inspectors questioned security personnel, reviewed the fire watch log and determined that the fire watch log had been checked off as completed. The completion time corresponded to the time the inspector was in the room. After further review and interviews with security personnel, the inspectors determined that the plant employee designated to perform the fire watch duties misunderstood the requirements for the fire watch. The employee had only verified the auxiliary building hallway area outside the room and failed to check inside the auxiliary building electrical penetration room as required. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-00869. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding had an adverse affect on the "Fixed Fire Protection Systems" element of fire watches posted as a compensatory measure for outages or degradations. The inspectors assigned a high degradation rating due to the fact that automatic fire suppression system was tagged out and inoperable. Because the system was degraded without compensatory actions for approximately 2 hours, the inspectors used a duration factor of 0.01. The inspectors used 2E-2 for a generic fire frequency area which corresponds to Table 1.4.2, Generic Fire Area Fire Frequencies for a switchgear room. The resulting calculated change in core damage frequency was 2E-4, which was greater than the high degradation Phase 1 Quantitative Screening Criteria of 1E-6, requiring a Phase 2 analysis. The inspectors consulted with a regional Senior Reactor Analyst and a simplified Phase 3 was performed using a duration factor of 2.3E-4 for the 2-hour time period, and the IPEEE specific room fire frequency of 7.2E-4. The resulting calculated change in core damage frequency was 1.7E-7, which would be less that the Phase 1 quantitative screening criteria. Using this information, the regional Senior Reactor Analyst, determined the finding to be of very low safety significance. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices in that the individual assigned to perform the fire watch proceeded in the face of uncertainty and failed to use appropriate human error prevention techniques. [H.4(a)] (Section 1R05) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure of Licensed Senior Reactor Operators to Maintain the Required Proficiency to Maintain Their License Current. The inspectors identified a noncited violation of 10 CFR 55.53.e, Conditions of License, for failure of licensed senior reactor operators to maintain the required proficiency to maintain their license current. Senior reactor operators standing the shift supervisor/shift technical advisor position were taking credit for senior reactor operator proficiency watches while standing this position. The normal shift complement of senior reactor operators consist of a shift manager, a control room supervisor, and a shift supervisor/shift technical advisor. When this issue was brought to the attention of operations management; they stopped the practice of the shift supervisor/shift technical advisor receiving senior reactor operator proficiency watch credit for standing that position. All shift supervisor/shift technical advisor senior reactor operators were inactivated. The plant issued a standing order that prohibited the shift supervisor/shift technical advisor to be allowed to perform the senior reactor operators oversight function in the control room and the shift manager or control room supervisor had to be in the control room at all times. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-01126. This finding was more than minor because if left uncorrected the finding could become a more significant safety concern. This finding affects the mitigating system cornerstone. The finding was determined to be of very low safety significance using the Licensed Operator Requalification Significance Determination Process since it related to operator license conditions and more than 20 percent of the affected individuals were deficient (Section 1R11). Inspection Report# : 2008002 (pdf)

Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform a Required Technical Specification Surveillance. The inspectors identified a noncited violation of Technical Specifications 3.8.1, AC Sources-Operating, for the failure to perform a required surveillance following the loss of a required offsite power source. The plant suffered a loss of power from the Port Gibson 115 kV line during high winds. Due to the fact that there is no direct control room alarm to alert the operating crew, they were not immediately aware they had lost the offsite source of power. When the crew recognized the loss of the bus they only entered a potential limiting condition of operations, due to the crew failing to realize that this was one of the required offsite sources. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2008-00737 and 2008-01202. This finding was more than minor because it impacts the mitigating system cornerstone objective in that it affects the operability, availability, reliability of an offsite power source that supplies a bus that provides power to mitigating systems. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was of very low safety significance since it did not represent an actual loss of a safety function. The cause of this finding has a crosscutting aspect in the area of human performance associated with the resources attribute in that the operators did not have adequate procedural guidance to determine the loss of safety-related offsite power supply. [H.2(c)] (Section 1R22) Inspection Report# : 2008002 (pdf) Significance: Mar 22, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate Cracks in Standby Service Water Pump House Structure. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, for failing to implement effective corrective actions after identifying concrete cracking in the standby service water pump houses. The inspectors determined that the program that evaluates, monitors, and repairs cracks for all safety related structures only identified a single crack for the entire site and does not track other structural cracks previously identified in the corrective action program. The last program inspection had been performed as recently as October 25, 2007, and only identified the single crack that had been documented in previous inspections. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2007-05824. This finding was more than minor because the cracks represent a degrading condition that if left uncorrected could become more significant safety concern. The inspectors determined this finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, this finding was of very low safety significance since it did not represent an actual loss of a safety function. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee personnel failed to properly maintain and utilize the program for evaluating, tracking and repairing identified concrete cracks in safety related structures. [H.4(b)] (Section 4OA2) Inspection Report# : 2008002 (pdf) Barrier Integrity Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct Leaking Reactor Water Cleanup System Primary Containment Isolation Valves The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, involving the failure to correct leaking reactor water cleanup system primary containment isolation valves. During refuelling Outage 16, plant personnel were performing local leak rate testing of reactor water cleanup backwash containment

penetration. Testing determined that these primary containment isolation valves exceeded the allowable leakage rate by greater than 10 times the leakage limits. The inspectors determined that for four consecutive operating cycles, the site had failed to take corrective actions to correct the excessive leakage through these valves. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 05139. The finding was more than minor because it was associated with systems, structures, and components and the reactor coolant system barrier performance attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the containment system. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee failed to take actions to correct a long-standing equipment issue associated with excessive leakage from primary containment isolation valves [H.2 (a)]. (Section 1R20) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Monitor Plant Parameters to Control Reactor Coolant System Cooldown Rate The inspectors identified a Green finding involving the failure to demonstrate proper monitoring of plant parameters to control reactor coolant system cooldown rate to within expected management standards. The plant experienced a reactor scram from approximately 15 percent power during plant start-up from a refuelling outage due to a total loss of feedwater. Reactor pressure decreased at a faster rate than expected due to low decay heat levels and the injection of relatively cold condensate storage tank water to reactor vessel. The control room supervisor did not give a pressure band after pressure decreased below the low end of the emergency operating procedure band of 800 psig or assign a licensed operator to monitor reactor pressure during the event. The inspectors identified to the operators that the plant was approaching the procedural limit for cooldown rate; operators then closed the inboard main steam isolation valves to prevent exceeding the cooldown rate. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 06201. The finding is more than minor since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not represent an actual degradation of the radiological barrier function of the reactor coolant system barrier. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because control room supervision failed to maintain proper oversight to ensure reactor coolant cooldown rate was maintained within procedural limits [H.1(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Jun 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Causing a Loss of Decay Heat Removal to the Spent Fuel Pool. The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure of operators to follow a safety-related off normal event procedure resulting in a loss of decay heat removal to the spent fuel pool. The operators elected to remove cooling to the fuel pool cooling heat exchangers to minimize the temperature rise on the component cooling water system during a partial loss of the plant service water system. This action was not specified in the off-normal event procedure. The off-normal event procedure only permitted the isolation of component cooling water flow to the fuel pool cooling heat exchangers for degraded component cooling water flow or pressure. This resulted in the spent fuel pool losing decay heat removal for approximately 3 hours and 22 minutes. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2008-02147.

The finding is more than minor since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not preclude operators from restoring spent fuel pool cooling to ensure the Fuel Barrier Cornerstone. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making in that operators did not use a systematic decision making process when faced with unexpected plant conditions [H.1(a)]. Inspection Report# : 2008003 (pdf) Significance: Mar 22, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Design Control of HPCS Minimum Flow Valve Motor-Operated Valve Over Current Setpoint. The inspectors identified a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to properly set the over current trip setpoint for the high pressure core spray minimum flow motor operated valve. This resulted in a spurious over current trip of the valve breaker during a high pressure core spray momentary pump start for breaker operability following post Division 3 emergency core cooling system testing. As a result of the trip, the high pressure core spray minimum flow valve failed open. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2008-01201. The finding was more than minor because it was associated with the barrier integrity cornerstone to provide reasonable assurance that the physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the MC 0609, "Significance Determination Process," Phase 1 worksheet, the finding was determined to have very low safety significance since it did not result in a loss of the containment barrier. Additionally, the issue was screened and determined to not impact the High Pressure Core Spray mitigating system function. (Section 4OA3) Inspection Report# : 2008002 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous

Last modified : April 07, 2009 Grand Gulf 1 1Q/2009 Plant Inspection Findings Initiating Events Significance: Feb 12, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Procedure Requirements for Preventive Maintenance Strategy Development A Green self revealing finding was identified for the failure to implement maintenance procedure requirements. Specifically, in June 2007, an incorrect preventive maintenance template was applied to the main transformer auxiliary power transfer switch resulting in a less than optimal preventive maintenance strategy. This was subsequently determined to be a contributing cause to the January 12th reactor scram. This issue is entered in the corrective action program as condition Report 2008 0174. The performance deficiency associated with this finding is the failure of maintenance and engineering personnel to implement the requirements of Procedure EN-DC-335, PM Basis Template, Section 5.2, PM Basis Template Development. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with work practices, in that the supervisory and management oversight of work activities were not employed such that nuclear safety was supported [H.4.(c)] (Section 4OA4). Inspection Report# : 2009007 (pdf) Significance: Feb 12, 2009 Identified By: NRC Item Type: FIN Finding Failure to Implement Procurement Engineering Procedure Requirements A Green self revealing finding was identified for the failure of engineering and maintenance personnel to implement procurement engineering procedure requirements. Specifically, in January, 2007 a procurement engineering evaluation determined that a difference in part numbers provided by a vendor was an administrative part number change. Consequently, a current transformer with a slightly different form, fit, and operating characteristic was installed in the generator/unit differential trip circuitry. This combined with other unknown circuit deficiencies and grid reactive load anomalies, resulted in a generator trip and reactor scram on March 21, 2008. The finding is entered in the corrective action program as Condition Report 2008-01476. The performance deficiency associated with this finding is the failure of procurement engineering personnel to implement the requirements of Procedure EN-DC-313, Procurement Engineering Process, Section 5.6, Administrative Part Number Changes, resulting in a less than optimal replacement part for a current transformer in the Unit/Generator differential trip circuitry. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with decision making in that procurement engineering did not use conservative assumptions and adopt a requirement to demonstrate a proposed action is safe to

proceed rather than to demonstrate that an action is unsafe to disprove the action [H.1.(b)] (Section 4OA4). Inspection Report# : 2009007 (pdf) Significance: Feb 12, 2009 Identified By: NRC Item Type: FIN Finding Failure to Implement Preventive Maintenance Procedure Requirements A Green self revealing finding was identified for the failure of to implement maintenance procedure requirements. Specifically, between 2002 and 2008, neither the preventive maintenance optimization program, nor the turbine 10-year plan prescribed a preventive maintenance strategy for the thyristor voltage regulator control portion of the main generator voltage regulating system. Consequently, on October 26, 2008, an under-excitation condition existed in the main generator following transfer from automatic to manual voltage regulator control, resulting in a generator and turbine trip and a reactor scram. The finding is entered in the corrective action program as Condition Report 2008-6241. The performance deficiency associated with this finding is the failure of maintenance and engineering personnel to implement the requirements of Procedure EN-DC-324, Preventive Maintenance Programs, Section 5.2, Process Overview, and Procedure EN-DC-335, PM Basis Template, Section 5.2, PM Basis Template Development. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with decision making, in that a systematic process was not employed for risk significant decision making and that roles and authority for decision making was not formally defined [H.1.(a)] (Section 4OA4). Inspection Report# : 2009007 (pdf) Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Trip of a Reactor Recirculation Pump During Pump Up-shift to Fast Speed Due to Ineffective Corrective Actions The inspectors reviewed a self-revealing Green finding involving a recirculation pump trip during pump up-shift to fast speed due to ineffective corrective actions. The plant had recently replaced the recirculation motor on Pump A during the refuelling outage and during investigation determined that the instantaneous over-current trip for the breaker had drifted low. The inspectors performed a review of condition reports and determined that reactor recirculation Pump B had tripped following motor replacement for the same reason in September 2007. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-06269. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not contribute to loss of function of mitigating equipment. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program in that the licensee failed to perform a thorough evaluation of a problem that resulted in a plant transient such that the resolution properly addressed the cause and extent of condition [P.1(c)]. (Section 1R20) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008

Identified By: Self-Revealing Item Type: FIN Finding Automatic Reactor Scram Caused by an Operator Inadvertently Closing Steam Supply Valves to the Reactor Feed Pump Turbine The inspectors reviewed a self-revealing Green finding involving an automatic reactor scram caused by an operator inadvertently closing steam supply valves to the reactor feed pump turbine. Site personnel investigating the scram determined that an operator had incorrectly performed actions for the reactor feed Pump B turbine on the reactor feed Pump A turbine control switches at a local panel. The operator inadvertently closed the steam supply valves to the reactor feed Pump A turbine resulting in a total loss of feedwater flow and low reactor water level scram. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-06195. The finding was more than minor because it was associated with the initiating events cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that an evaluation was required by the regional senior reactor analyst, because the finding impacted both the initiating event and mitigating systems cornerstone. The senior reactor analyst performed a Phase 3 analysis and determined the issue was very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the operator failed to use proper self-checking techniques while performing actions to place feed Pump B in the standby lineup [H.4(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Jun 21, 2008 Identified By: Self-Revealing Item Type: FIN Finding Ineffective Corrective Actions in Response to Plant Transients Resulting from Animal Intrusions. The inspectors reviewed a self-revealing Green finding involving ineffective corrective actions that resulted in an unplanned down power caused by an animal intrusion. The plant experienced a loss of the balance of plant Transformer 23 with a loss of power to the plant service water pumps. Operators reduced reactor power to 47 percent. The control room dispatched operators to the river via a boat due to high river level and discovered a dead raccoon in the vicinity of the transformer. The inspectors noted that two previous reactor scrams had been caused by raccoons, and an injured raccoon had previously been found at the base of Transformer 23. The inspectors concluded that the flooding conditions which have been routinely experienced at the site and the occurrence of raccoon events at the site could have been used to anticipate and mitigate the unplanned down power. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2008-02089. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because the finding impacted both the Initiating Event and Mitigating Systems Cornerstone. The inspectors performed a Phase 2 analysis using Appendix A Determining the Significance of Reactor Inspection Findings for At-Power Situations, of Manual Chapter 0609, Significance Determination Process, and the Phase 2 Worksheets for Grand Gulf Nuclear Station. The inspectors determined there was an increase in likelihood of a transient without the power conversion system but there was no reduction in remaining capability. Because the exposure time of the finding was less than 30 days, the result of the Phase 2 analysis was that the finding had very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with operating experience in that the licensee failed to implement proper corrective actions to prevent animals from causing a plant transient [P.2(b)]. Inspection Report# : 2008003 (pdf) Mitigating Systems

Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedures to Maintain Drains on Safety Related Buildings The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, involving the failure to properly clean and inspect the rooftop and associated water drainage systems of the safety-related diesel generator building. The inspectors identified loose, flexible roofing material that could have covered roof drains and result in loss of functionality for all of the standby diesel generators during a design basis heavy rainfall event. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-00429. This finding is more than minor since it affects the protection against external events attribute of mitigating system cornerstone. The roofing material and debris represented a degrading condition that if left uncorrected could have affected the availability, reliability, and capability of the standby diesel generators to respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding screened as potentially risk significant due to a flooding or severe weather initiating event, which then required a Phase 3 analysis. The Phase 3 analysis calculated a change in core damage frequency of 3.04E-8/yr, which represented very low safety significance. Inspection Report# : 2009002 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation for Standby Service Water Cooling Tower Drift Eliminators The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, involving a failure to perform an adequate operability evaluation. The inspectors identified non-conservatisms in the evaluation with regards to standby service water cooling tower drift rate, a failure to consider external events design basis impacts, and a failure to properly classify the condition as a substantially degraded, non-conforming condition, because it was subsequently determined that the deficiency could increase drift losses by a factor of ten. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-01222. This finding is more than minor because the failure to perform adequate operability evaluations, if left uncorrected, could become a more significant safety concern because the loss rates could become worse over time. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions. Inspection Report# : 2009002 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: FIN Finding Inadequate Fire Drill Critique The inspectors identified a finding for fire brigade performance deficiencies that were not identified by the licensee during a fire drill critique. The inspectors identified several deficiencies during the drill including issues relating to command and control, fire fighting strategy and use of fire fighting equipment. The inspectors provided feedback to plant personnel on the identified performance issues and the inadequate drill evaluation. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 06522. This finding was more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone objective and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, was used to analyze the finding since the inadequate critique had an adverse effect on fire brigade effectiveness, in relation to defense-in-depth

strategies. Manual Chapter 0609, Appendix F states that findings associated with the onsite manual fire brigade are excluded. Therefore, in accordance with Manual Chapter 0609, the safety significance was determined by regional management review. Regional management concluded that the finding was of very low safety significance because it reflected fire brigade performance during a training drill, rather than during an actual fire. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to have a low enough threshold in identifying performance issues associated with a plant fire drill [P.1(a)]. (Section 1R05) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of the Engineering Safety Features Electrical Switchgear and Battery Room Ventilation System The inspectors identified a noncited violation of 10 CFR 50.65(a)(2) for the failure to adequately monitor the performance of the engineering safety features electrical switchgear and battery room ventilation system. The inspectors identified a condition report from March 2004 that had not been screened and evaluated in the maintenance rule database as a maintenance preventable functional failure. The condition report identified a room cooler that had tripped due to excessive current on the fan motor because an incorrectly sized sheave was installed during previous maintenance. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 02219. The inspectors determined that this finding was more than minor since the engineering safety features electrical switchgear and battery room ventilation system was not placed in (a)(1) monitoring status in a timely manner. In addition, the finding was more than minor since violations of 10 CFR 50.65(a)(2) necessarily involve degraded system performance, which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the maintenance rule aspect of the finding did not lead to an actual loss of safety function of the system or cause a component to be inoperable, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. (Section 1R12) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of Inadequate Operability Evaluations The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving two examples of a failure to follow procedures which resulted in inadequate operability evaluations. The first example involved an inadequate evaluation of foreign material in the condensate storage tank. The evaluation relied on an assumption that the high-pressure core spray and reactor core isolation cooling pumps would not be damaged by metal debris entrained in the pumps suction. The second example involved an inadequate evaluation of the structural integrity of the standby service water cooling towers that only considered the loss of structural support from a single beam. The licensee entered these issues into the corrective action program as Condition Reports CR GGN 2008 05685 and CR GGN 2008 06044. This finding is more than minor because the failure to perform adequate operability evaluations, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions [H.1(b)]. (Section 1R15 Inspection Report# : 2008005 (pdf)

Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Standby Service Water Corrosion The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving a failure to take corrective actions to prevent recurrence of severe corrosion in piping hangers, piping supports, and piping in the standby service water basin cooling towers. Significant corrosion of the standby service water supports in October 2008 had been previously identified by plant personnel during a ten-year in-service inspection on October 3, 1993. At that time, plant personnel determined this to be a significant degraded condition of a safety related system, requiring replacement of the piping and associated supports. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-05434. This finding was more than minor because the corrosion represented a degrading condition that if left uncorrected could become more significant safety concern. The finding was also more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not represent an actual loss of safety function, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Following Identification of Degrading Standby Service Water Supports The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, involving the failure to take timely corrective actions for corrosion on distribution beam structural support posts in the standby service water basin cooling towers. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-05434. The finding was more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance since it did not represent an actual loss of safety function of the standby service water cooling towers, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because licensee personnel failed to identify issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Nov 06, 2008 Identified By: NRC Item Type: NCV NonCited Violation B.5.b. Phase 2 and 3 Mitigating Strategy This finding, affecting the Mitigating Systems Cornerstone, is related to mitigative measures developed to cope with losses of large areas of the plant; in response to Section B.5.b. of the February 25, 2002, Interim Compensatory Measures (ICM) Order (EA-02-026) and related NRC guidance. This finding has been designated as "Official Use Only - Security-Related Information;" therefore, the details of this finding are being withheld from public disclosure. This finding has no cross-cutting aspect. See inspection report 2008-007 for more details. Inspection Report# : 2008007 (pdf)

Significance: Sep 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of Four Maintenance Rule Systems. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of four Maintenance Rule systems. Several discrepancies in the Maintenance Rule Program were discovered by the inspectors, including unevaluated condition monitoring failures in the neutron monitoring system and an unevaluated functional failure in the standby gas treatment system. Plant personnel implemented additional corrective actions to fully investigate the potential extent of this condition and the apparent weakness in the condition report screening process used for the Maintenance Rule program. As a result, the Maintenance Rule expert panel classified four systems as needing increased monitoring and goal setting, moving these systems from an a(2) to an a(1) status. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-02219. This finding is more than minor since it was similar to Inspection Manual Chapter 0612, Appendix E, Example 7.b in that the problem involved degraded equipment performance. This finding was characterized under the significance determination process as having very low safety significance because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel failed to use proper self-checking and peer-checking to identify repetitive maintenance rule functional failures and also failed to properly document condition report screening activities [H.4(a)] (Section 1R12). Inspection Report# : 2008004 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment during Adverse Weather Conditions. The inspectors identified a Green noncited violation of 10 CFR 50.65 (a)(4), involving the failure to perform risk assessments following multiple declared tornado watches affecting Grand Gulf Nuclear Station during the landfall of Hurricane Gustav. On the morning of September 3, 2008, the inspectors noted that the licensee had not evaluated the increased risk from a declared tornado watch for the Claiborne County area. The inspectors brought this to the attention of plant personnel and a risk assessment was performed and plant risk was changed from a Green to a Yellow risk condition. The inspectors then reviewed the tornado watches declared by the National Weather Service that affected Claiborne County during the landfall of Hurricane Gustav, and noted that six separate tornado watches had been declared over the previous three days. A review of the control room logs showed no documentation of changes in plant risk condition. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-04397. This finding is more than minor because the risk assessments failed to consider unusual external conditions that were present or imminent. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, Assessment of Risk Deficit and consulting with the regional senior risk analyst, the inspectors determined the finding of very low safety significance due to a calculated incremental core damage probability deficit of 4.38E-08. This finding has a crosscutting aspect in the area of human performance associated with work practices in that plant personnel failed to follow the risk management procedure [H.4(b)] (Section 1R13). Inspection Report# : 2008004 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: VIO Violation Failure to Perform an Adequate Inspection of PMP Door Seals Protecting Safety Related Equipment. The inspectors identified a Green cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. The licensee had previously received a noncited violation

for inadequate inspections of probable maximum precipitation door seals in NRC Inspection Report 05000416/2008002. On July 9, 2008, the inspectors found the entrance door to the Train B standby service water pump house not meeting the standards of the maintenance procedure because the door seals failed to make contact with the door. The extent of condition review found seven additional door seals degraded, including the doors to the diesel generator building and control building. The door seal on the Train B standby service water pump house identified by the inspectors on July 9, 2008, had not been identified by plant personnel during an extent of condition review on February 29, 2008. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-03216. The finding is more than minor since it affects the protection against external factors attribute of mitigating system cornerstone. The door seals also represent a degrading condition that if left uncorrected could affect the availability, reliability, and capability of mitigating systems required to respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multi-trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis using many bounding and conservative assumptions. The result was a delta-CDF of 3.3E 7/yr and a delta-LERF of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to take adequate corrective actions to ensure degraded probable maximum precipitation door seals were properly evaluated and repaired in a timely manner [P.1(d)] (Section 4OA2). Inspection Report# : 2008004 (pdf) Inspection Report# : 2009002 (pdf) Significance: Jun 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Fireproofing on Fire Barrier Protecting the Safeguards Switchgear Room The inspectors identified a Green noncited violation of Facility Operating License Condition 2.C(41) involving the failure to ensure that fire barriers protecting safety-related areas were functional. The inspectors identified an 8-foot length of structural steel in the east stairwell wall, which is shared by the Division I safeguards switchgear room, that did not have the required fireproofing to maintain an adequate fire barrier. The missing passive fire protection reduced the fire rating of the wall by allowing heat to transfer through the unprotected steel, thus degrading the fire containment capability assumed in the fire hazards analysis. The licensee entered this issue into their corrective action program as Condition eport CR GGN 2008 01849. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire confinement category. The inspectors assigned a high degradation rating due to the fact that the required fireproofing was missing. The inspectors used the supplemental screening process for fire confinement findings and concluded that the finding was of very low safety significance (Green) due to the fact that the degraded barrier would have provided a minimum of 20 minutes fire endurance protection and there were no fire ignition sources or combustible materials in the area that would subject the barrier to direct flame impingement. Inspection Report# : 2008003 (pdf) Significance: Jun 21, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Recognize the Division III Diesel Generator being Non-Functional. The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1(a) involving the failure to follow a system operating instruction. While shutting down the Division III diesel generator, operators failed to place the outside air fan in automatic alignment resulting in the Division III diesel generator being nonfunctional. On May 5, 2008, operators had shutdown the Division III diesel generator, but they failed to recognize that the outside

air fan was not running when they depressed the shutdown pushbutton for the outside air fan per the system operating instruction. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-02265 The finding is more than minor since it affects configuration control attribute of the Mitigating System Cornerstone objective, in that it affected the availability, reliability and capability of an onsite power source that supplies a bus that provides power to mitigating systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not represent a loss of a safety function that exceeded the Technical Specification allowed outage time. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices in that the operating crew did not use the proper human performance techniques of self checking while securing the outside air fan for the Division III diesel generator [H.4(a)]. Inspection Report# : 2008003 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate Fire Wrap Testing Discrepancies A noncited violation of License Condition 2.C(41), "Fire Protection Program," was identified because the licensee failed to evaluate vendor fire test results to ensure that deviations from the acceptance criteria did not result in a reduction in the effectiveness of the approved Fire Protection Program. The licensee replaced existing fire barrier material installed on conduits with 3M Interam fire wrap without recognizing that applicable NRC test criteria were not met. As a result, the licensee failed to perform an evaluation to determine whether the test results would result in a reduction in the effectiveness of the fire protection provided to the cables inside the affected conduits. The new fire wrap was installed to protect redundant trains of cables necessary for safe shutdown between 2004 and 2007. This finding was entered into the licensees corrective action program under Condition Report 2008-01910. The licensee took prompt compensatory measures and implemented hourly fire watches while the issue was being evaluated. Failure to properly evaluate vendor fire test results that did not satisfy the acceptance criteria in Generic Letter 86-10, Supplement 1 prior to changing the existing fire wrap with 3M Interam fire wrap as required by the approved Fire Protection Program was a performance deficiency. This finding was more than minor because it affected the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone Objective to ensure the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. This performance deficiency was also similar to the "more than minor" portion of Inspection Manual Chapter 0612, Appendix B, Example 3.i, in that an engineering evaluation was necessary to determine the acceptability of the existing fire wrap to perform its intended function. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected fire protection defense-in-depth strategies involving post-fire safe shutdown systems. This finding screened as having very low safety significance because it involved a fire barrier with a low degradation, since the nonconforming condition was subsequently determined to provide an acceptable margin to damage for the cables being protected. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Fire Brigade Members Assigned Responsibilities That Conflicted with Fire Brigade Responsibilities. A noncited violation of License Condition 2.C.(41) was identified for failure to maintain required staffing available to respond to a fire. Specifically, the approved Fire Protection Program requires that a five-person fire brigade be available onsite at all times and not assigned duties that conflict with the duties of the fire brigade. Contrary to this, on three occasions in March 2008, operators assigned as fire brigade members were directed to perform operator rounds at the radial wells. Because the Mississippi River was at flood stage, this required traveling by boat, so the operators were unable to return to the plant promptly for approximately 2 hours. This was further complicated by the fact that operator/fire brigade radios did not work during most of the boat trip and in the vicinity of the most distant well, meaning that operators could not be quickly recalled. This finding was entered into the corrective action program

under Condition Report 2008-01616. This finding had a crosscutting aspect in the area of Human Performance - Work Control (H.3.b) because the licensee did not ensure that different job activities were coordinated to ensure that the fire brigade remained available at all times. Failure to maintain a fully staffed fire brigade available onsite at all times was a performance deficiency. This finding was more than minor because it affected the protection from external factors (fire) attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesireable consequences. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected a fire protection defense-in-depth element. This finding was assigned a low degradation rating because the operations shift during the times when the fire brigade member was unavailable included extra fire brigade-trained personnel that could supplement the fire brigade. The delay in a replacement person reporting to the scene of a fire would not have impacted the initial fire fighting effort, since enough fire brigade personnel were available to perform the functions. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Plant Modification Blocked Access for Manual Firefighting. A noncited violation of License Condition 2.C.(41), "Fire Protection Program," was identified related to making a plant change that negatively impacted the effectiveness of the approved Fire Protection Program. The team identified that the licensee had permanently blocked the door to the abandoned Unit 2 portion of the joint control room without performing a fire protection impact evaluation. The only remaining access path was a small hatch that would have made it difficult for fire fighters to gain access with protective clothing and equipment. A fire in this area could threaten operations in the Unit 1 control room if not promptly suppressed. This finding was determined to have a cross-cutting aspect in problem identification and resolution timeliness (P.1.d) because fire protection personnel recognized that a new access door was needed in 2006, but no substantial action had been taken to install it by the time of this inspection. This finding was entered in to the licensees corrective action program under Condition Reports 2008-001893 and 2008-01913. Blocking access to the Unit 2 control room area and not promptly restoring access to allow manual fire suppression was a performance deficiency. This finding was more than minor because it affected the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events in order to prevent undesirable consequences. This finding was evaluated using Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," because it affected a fire protection defense-in-depth element. This finding was determined to have very low safety significance because all potential fire ignition sources in the affected area screened out in Task 2.3.4 in the Phase 2 evaluation. There were no ignition sources because the licensee had removed electrical power from this area, and administratively prevented hot work and storage of transient combustible material from this area. Inspection Report# : 2008006 (pdf) Significance: Apr 18, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure That Potential Damage to Motor-Operated Valve Circuits Would Not Prevent Safe Shutdown. A noncited violation of 10 CFR Part 50, Appendix R, Section III.G.1.a was identified because the licensee failed to evaluate the impact of a potential motor operated valve failure mechanism on the ability to implement post-fire safe shutdown following a control room evacuation. The team identified that the Residual Heat Removal Pump Minimum Flow Valve F064A could be damaged by fire in the control room and not be available to perform its safe shutdown function. This finding involved mechanistic damage due to hot shorts as described in Information Notice 92-18, "Potential for Loss of Remote Shutdown Capability During Control Room Fire." The licensee had incorrectly interpreted this operating experience and concluded that no action was required. This finding was entered into the corrective action program under Condition Reports 1999-0236 and 2008-01904.

The team determined that failure to ensure that components necessary to safely shutdown the reactor would remain operable following a fire was a performance deficiency. This deficiency was more than minor because it impacted the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to external events (fire) to prevent undesirable consequences. The Phase 3 risk evaluation performed by the senior reactor analyst determined this deficiency had very low safety significance because the probability of having a fire in either of the two control room panels where the postulated damage could occur and lead to a control room evacuation was very low. Inspection Report# : 2008006 (pdf) Barrier Integrity Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Enter a Limiting Condition for Operation for Primary Containment Isolation Valves The inspectors identified a Green noncited violation of Grand Gulf Nuclear Station Technical Specifications 3.6.1.3, for failure to enter a limiting condition for operation action statement for primary containment isolation valves. As a result, the limiting condition for operation action statement time was exceeded. The inspectors identified that surveillance test data for the residual heat removal Train A minimum flow valve was missing. The inspectors discovered that operations staff failed to properly review the work order for the valve work, and they had made an assumption the work order had been canceled. The licensee reviewed the identified issue for extent of condition and identified that in addition to a missed postmaintenance stroke test, they had also failed to enter the limiting condition for operation for two containment isolation valves. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-01069. This finding was more than minor since it affects the configuration control attribute of barrier integrity cornerstone, in that failing to properly test containment isolation valves could affect the assurance that physical design barriers that protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the containment system. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices, in that the operations shift supervisor and maintenance coordinator failed to perform proper self- and peer-checking and proper documentation of the completed work activity. Inspection Report# : 2009002 (pdf) Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Correct Leaking Reactor Water Cleanup System Primary Containment Isolation Valves The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, involving the failure to correct leaking reactor water cleanup system primary containment isolation valves. During refuelling Outage 16, plant personnel were performing local leak rate testing of reactor water cleanup backwash containment penetration. Testing determined that these primary containment isolation valves exceeded the allowable leakage rate by greater than 10 times the leakage limits. The inspectors determined that for four consecutive operating cycles, the site had failed to take corrective actions to correct the excessive leakage through these valves. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 05139. The finding was more than minor because it was associated with systems, structures, and components and the reactor coolant system barrier performance attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an

actual open pathway in the physical integrity of the containment system. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee failed to take actions to correct a long-standing equipment issue associated with excessive leakage from primary containment isolation valves [H.2 (a)]. (Section 1R20) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Monitor Plant Parameters to Control Reactor Coolant System Cooldown Rate The inspectors identified a Green finding involving the failure to demonstrate proper monitoring of plant parameters to control reactor coolant system cooldown rate to within expected management standards. The plant experienced a reactor scram from approximately 15 percent power during plant start-up from a refuelling outage due to a total loss of feedwater. Reactor pressure decreased at a faster rate than expected due to low decay heat levels and the injection of relatively cold condensate storage tank water to reactor vessel. The control room supervisor did not give a pressure band after pressure decreased below the low end of the emergency operating procedure band of 800 psig or assign a licensed operator to monitor reactor pressure during the event. The inspectors identified to the operators that the plant was approaching the procedural limit for cooldown rate; operators then closed the inboard main steam isolation valves to prevent exceeding the cooldown rate. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 06201. The finding is more than minor since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not represent an actual degradation of the radiological barrier function of the reactor coolant system barrier. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because control room supervision failed to maintain proper oversight to ensure reactor coolant cooldown rate was maintained within procedural limits [H.1(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Jun 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedures Causing a Loss of Decay Heat Removal to the Spent Fuel Pool. The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) involving the failure of operators to follow a safety-related off normal event procedure resulting in a loss of decay heat removal to the spent fuel pool. The operators elected to remove cooling to the fuel pool cooling heat exchangers to minimize the temperature rise on the component cooling water system during a partial loss of the plant service water system. This action was not specified in the off-normal event procedure. The off-normal event procedure only permitted the isolation of component cooling water flow to the fuel pool cooling heat exchangers for degraded component cooling water flow or pressure. This resulted in the spent fuel pool losing decay heat removal for approximately 3 hours and 22 minutes. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2008-02147. The finding is more than minor since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not preclude operators from restoring spent fuel pool cooling to ensure the Fuel Barrier Cornerstone. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making in that operators did not use a systematic decision making process when faced with unexpected plant conditions [H.1(a)].

Inspection Report# : 2008003 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : June 05, 2009

Grand Gulf 1 2Q/2009 Plant Inspection Findings Initiating Events Significance: Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Respond to Control Room Alarms in Accordance with Plant Procedures A self-revealing noncited violation of Technical Specification 5.4.1(a) was reviewed involving a failure to follow the fire alarm response procedure during a fire in the reactor feedwater pump area. The operators failed to investigate the source of a smoke alarm for an hour, allowing a fire to develop beyond the incipient stage before it was discovered. On November 17, 2008, a fire ignited in oil-soaked insulation on the reactor feedwater Pump B. After two weeks of plant operation following a refueling outage, during the November 17 shift turnover meeting, a fire alarm was received in the control room and was acknowledged by an operator. No notifications were made to the shift manager, and no operator or fire brigade member was dispatched. One hour after shift turnover, during normal operator rounds the turbine building operator discovered the fire in the reactor feedwater pump room. The fire brigade was dispatched to extinguish the fire. The licensee entered this condition in the corrective action program as condition report CR-GGN-2008-06584. The finding was more than minor because it was associated with the initiating events cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The finding was assessed by performing a bounding analysis using Appendix M of Inspection Manual Chapter 0609. The bounding analysis indicated that the change in core damage frequency would be 4.24 x 10-7 over a 1-year assessment period, indicating that this finding was of very low safety significance. This finding has a crosscutting aspect in the area of human performance with a work practices component for failure to use proper self checking techniques commensurate with the risk of the assigned task to ensure the work is performed safely because operators failed to use self checking techniques when acknowledging the reactor feedwater pump fire alarm [H.4(a)] (Section 4OA3). Inspection Report# : 2009003 (pdf) Significance: Feb 12, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Procedure Requirements for Preventive Maintenance Strategy Development A Green self revealing finding was identified for the failure to implement maintenance procedure requirements. Specifically, in June 2007, an incorrect preventive maintenance template was applied to the main transformer auxiliary power transfer switch resulting in a less than optimal preventive maintenance strategy. This was subsequently determined to be a contributing cause to the January 12th reactor scram. This issue is entered in the corrective action program as condition Report 2008 0174. The performance deficiency associated with this finding is the failure of maintenance and engineering personnel to implement the requirements of Procedure EN-DC-335, PM Basis Template, Section 5.2, PM Basis Template Development. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with work practices, in that the supervisory and management oversight of work activities were

not employed such that nuclear safety was supported [H.4.(c)] (Section 4OA4). Inspection Report# : 2009007 (pdf) Significance: Feb 12, 2009 Identified By: NRC Item Type: FIN Finding Failure to Implement Procurement Engineering Procedure Requirements A Green self revealing finding was identified for the failure of engineering and maintenance personnel to implement procurement engineering procedure requirements. Specifically, in January, 2007 a procurement engineering evaluation determined that a difference in part numbers provided by a vendor was an administrative part number change. Consequently, a current transformer with a slightly different form, fit, and operating characteristic was installed in the generator/unit differential trip circuitry. This combined with other unknown circuit deficiencies and grid reactive load anomalies, resulted in a generator trip and reactor scram on March 21, 2008. The finding is entered in the corrective action program as Condition Report 2008-01476. The performance deficiency associated with this finding is the failure of procurement engineering personnel to implement the requirements of Procedure EN-DC-313, Procurement Engineering Process, Section 5.6, Administrative Part Number Changes, resulting in a less than optimal replacement part for a current transformer in the Unit/Generator differential trip circuitry. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with decision making in that procurement engineering did not use conservative assumptions and adopt a requirement to demonstrate a proposed action is safe to proceed rather than to demonstrate that an action is unsafe to disprove the action [H.1.(b)] (Section 4OA4). Inspection Report# : 2009007 (pdf) Significance: Feb 12, 2009 Identified By: NRC Item Type: FIN Finding Failure to Implement Preventive Maintenance Procedure Requirements A Green self revealing finding was identified for the failure of to implement maintenance procedure requirements. Specifically, between 2002 and 2008, neither the preventive maintenance optimization program, nor the turbine 10-year plan prescribed a preventive maintenance strategy for the thyristor voltage regulator control portion of the main generator voltage regulating system. Consequently, on October 26, 2008, an under-excitation condition existed in the main generator following transfer from automatic to manual voltage regulator control, resulting in a generator and turbine trip and a reactor scram. The finding is entered in the corrective action program as Condition Report 2008-6241. The performance deficiency associated with this finding is the failure of maintenance and engineering personnel to implement the requirements of Procedure EN-DC-324, Preventive Maintenance Programs, Section 5.2, Process Overview, and Procedure EN-DC-335, PM Basis Template, Section 5.2, PM Basis Template Development. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with decision making, in that a systematic process was not employed for risk significant decision making and that roles and authority for decision making was not formally defined [H.1.(a)] (Section 4OA4).

Inspection Report# : 2009007 (pdf) Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Trip of a Reactor Recirculation Pump During Pump Up-shift to Fast Speed Due to Ineffective Corrective Actions The inspectors reviewed a self-revealing Green finding involving a recirculation pump trip during pump up-shift to fast speed due to ineffective corrective actions. The plant had recently replaced the recirculation motor on Pump A during the refuelling outage and during investigation determined that the instantaneous over-current trip for the breaker had drifted low. The inspectors performed a review of condition reports and determined that reactor recirculation Pump B had tripped following motor replacement for the same reason in September 2007. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-06269. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not contribute to loss of function of mitigating equipment. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program in that the licensee failed to perform a thorough evaluation of a problem that resulted in a plant transient such that the resolution properly addressed the cause and extent of condition [P.1(c)]. (Section 1R20) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Automatic Reactor Scram Caused by an Operator Inadvertently Closing Steam Supply Valves to the Reactor Feed Pump Turbine The inspectors reviewed a self-revealing Green finding involving an automatic reactor scram caused by an operator inadvertently closing steam supply valves to the reactor feed pump turbine. Site personnel investigating the scram determined that an operator had incorrectly performed actions for the reactor feed Pump B turbine on the reactor feed Pump A turbine control switches at a local panel. The operator inadvertently closed the steam supply valves to the reactor feed Pump A turbine resulting in a total loss of feedwater flow and low reactor water level scram. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-06195. The finding was more than minor because it was associated with the initiating events cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that an evaluation was required by the regional senior reactor analyst, because the finding impacted both the initiating event and mitigating systems cornerstone. The senior reactor analyst performed a Phase 3 analysis and determined the issue was very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the operator failed to use proper self-checking techniques while performing actions to place feed Pump B in the standby lineup [H.4(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Mitigating Systems Significance: Jun 23, 2009

Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Design Changes to Protect the Standby Service Water Slab The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III involving the failure to incorporate design changes required to limit dynamic loads on the standby service water basin slab. In 1997, the plant experienced damage to the standby service water basin slabs resulting from unanalyzed dynamic loads. During a standby service water system inspection on April 18, 2009, inspectors observed several different tire tracks on the seismically-designed concrete slab that covers and is integral to the safety-related standby service water basin. The inspectors also noted small placards attached to the basin slabs which prohibited moving vehicles on the slabs, and other signs requiring protective mats under any items placed on the slabs. Plant personnel evaluated the loading of the vehicle and determined that the load limits on the basin slab had not been exceeded. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2009 002087. The inspectors determined this finding affected the design control 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. In addition, the finding was more than minor because the failure to prevent dynamic loads on the standby service water basin slabs, if left uncorrected, could become more significant safety concern. Using the Manual Chapter of 0609, Significance Determination Process, Phase 1 Worksheet, this finding was determined to have very low safety significance, because it did not represent an actual loss of a safety function of the standby service water system. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue occurred several years ago and does not reflect current licensee performance (Section 1R04). Inspection Report# : 2009003 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation for Debris Left in the Condensate Storage Tank The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving a failure to follow procedures which resulted in an inadequate operability evaluation. During the week of May 18, 2009, the site conducted debris removal in the condensate storage tank. This debris removal was necessary because of a failure to remove all debris in the condensate storage tank during their spring 2007 cleanup project. The licensee performed an operability evaluation for objects left in the condensate storage tank which stated that the high pressure core spray system and reactor core isolation cooling would remain operable for all postulated events. Upon review by the inspectors, the operability evaluation did not address several issues including objects left in the condensate storage tank and condensate system return flow to the condensate storage tank following a plant shutdown/scram. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2009-02815 and CR-GGN-2009-02837. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to thoroughly identify all materials left in the condensate storage tank during their original operability determination [P.1(a)] (Section 1R15). Inspection Report# : 2009003 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: FIN Finding Failure to Perform a Timely Operability Evaluation Following the Discovery of a SSW Fan Failure Mechanism The inspectors identified a Green finding involving the failure to perform an operability determination after a new failure mechanism was discovered for standby service water Fan B. The inspectors were performing a follow up

review of a condition report that detailed a trip of Division 1 standby service water Fan B. The fan had tripped on start up from the control room on December 31, 2007. The licensee had initially determined the trip was due to a faulty solid state trip device. Subsequent testing in failed to identify a problem with the trip device, and the apparent cause of the fan trip was attributed to reverse rotation of the fan. Operations personnel were not informed of this new information as required by the corrective action program procedure. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2009-01726. This finding is more than minor because it was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that the finding had very low safety significance (Green) since it did not represent an actual loss of a safety function of the standby service water cooling towers, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of human performance associated with work practices in that licensee personnel failed to apply procedural requirements to write a new condition report when new information was acquired related to the trip of the Division 1 standby service water Fan B [H.4(b)] (Section 4OA2). Inspection Report# : 2009003 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of Division II Standby Diesel Generator Jacket Water Heat Exchanger The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," for failure to comply with the licensees Generic Letter 89 13 program, which specifically states that cleaning of heat exchangers covered by this program is prohibited prior to performing an as-found thermal performance test. Specifically, in early 2006, the Division II Standby Diesel Generator (i.e. Emergency Diesel Generator) jacket water cooling heat exchanger was cleaned just prior to performing a five year thermal performance test. The licensee has entered this into their corrective action program as CR-GGN-2009-00904. This finding is more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of safety systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The inspectors reviewed the finding for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Bias in Instrumentation Used for Standby Service Water Leak Detection The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to establish adequate measures for the selection and review for suitability of equipment and processes that are essential to the safety-related functions of structures, systems and components. Specifically, the licensee failed to properly design for pulsation effects on flow rate instrumentation used for leak detection in the Standby Service Water system. This instrumentation is needed to meet licensee commitment 10 CFR Part 50, Appendix A, General Design Criterion 13, "Instrumentation and Control," to monitor trends in the ultimate heat sink basin inventory with the system in operation. The system was designed to detect a leakage rate of 1250 gallons per minute, and alarm in the control room at this leak rate, but due to design inadequacies in the instrumentation, the leak rate would have to exceed 3350

gallons per minute before activating the alarm. The licensee has entered this into their corrective action program as CR GGN 2009-00054. This finding was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1 3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The finding was reviewed for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Motor-Operated Valve Calculations Used Non-Conservative Inputs and Methodologies The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" for failing to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee used non conservative inputs or methodologies in calculating terminal voltages to safety related motor-operated valve motors that would be required to operate for mitigation of design bases events. The licensees electrical calculations used non conservative 50 percent locked rotor currents and neglected thermal overload resistance to determine the terminal voltages to safety related motor-operated valves which would predict higher terminal voltages than would actually exist. The calculated terminal voltages were direct design inputs into the applicable motor-operated valves mechanical thrust and torque calculations. The licensee has entered this issue into their corrective action program as CR GGN 2009 00985. This finding was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1 3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. This finding has a cross cutting aspect in the area of Problem Identification and Resolution, in that self assessments are of sufficient depth, are comprehensive, are appropriately objective, and are self critical. The licensee had conducted a Component Design Bases Assessment, LO GLO 2008 00044 in August 2008, and failed to adequately assess an identical finding identified at River Bend Station during their 2008 Component Design Bases Inspection. The licensee had determined that this issue was not applicable at Grand Gulf Nuclear Station. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions fo rReplacement of Safety-Related Batteries The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for failure to identify and correct a condition adverse to quality related to the seismic qualification of the Division III High Pressure Core Spray safety-related battery. Specifically, the licensee failed to identify an incorrectly installed end bracket after replacement of the Division III safety-related battery in 2002 using procedures, work instructions, and drawings that were supposed to have been corrected after this same issue was identified during a 1997 battery replacement activity. The licensee has entered this into their corrective action program as CR-GGN-2009-00830. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events

for ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was confirmed to not result in a loss of operability or functionality. The finding was reviewed for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of a Failure to Meet 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," with two examples. Specifically, the team identified that the licensee failed to develop and implement adequate testing programs for Class 1E molded-case circuit breakers, and for the voltage and frequency response of the standby diesel generators that met design or vendor requirements and recommendations. In response, the licensee entered these examples in the corrective action program as CR GGN 2009 01024, and CR GGN-2009-01057. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events for ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, each example was determined to be of very low safety significance (Green) because they did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding has a cross cutting aspect in the area of Problem Identification and Resolution, in that self assessments are of sufficient depth, are comprehensive, are appropriately objective, and are self critical. The licensee had conducted a Component Design Bases Assessment, LO-GLO-2008-00044 in August 2008, and failed to adequately assess an identical finding identified at River Bend Station during their 2008 Component Design Bases Inspection. The licensee had determined that this issue was not applicable at Grand Gulf Nuclear Station. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control for Standby Service Water Pump Cables and Electrical Vaults The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to adequately demonstrate operability for the 4160 volt Standby Service Water Pump kerite cables through adequate testing and analysis in a continuously submerged environment. Furthermore, the environment for these continuously submerged cables exists because each of the two vaults that contain these cables (MH 20 and MH 21) has a design flaw, in that several other vaults gravity drain to them and the design of these vaults did not include a sump pump or other means for water to be removed or drained from them. The licensee has entered this into their corrective action program as CR-GGN-2009-01028. This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control of ensuring the availability, reliability, and capability of safety systems, and closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, because there was reasonable doubt on the continued operability of the Standby Service Water system. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution in that the licensee failed to implement Operating Experience directly communicated with a Generic Letter through changes to station processes, procedures, and equipment.

Inspection Report# : 2009006 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedures to Maintain Drains on Safety Related Buildings The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, involving the failure to properly clean and inspect the rooftop and associated water drainage systems of the safety-related diesel generator building. The inspectors identified loose, flexible roofing material that could have covered roof drains and result in loss of functionality for all of the standby diesel generators during a design basis heavy rainfall event. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-00429. This finding is more than minor since it affects the protection against external events attribute of mitigating system cornerstone. The roofing material and debris represented a degrading condition that if left uncorrected could have affected the availability, reliability, and capability of the standby diesel generators to respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding screened as potentially risk significant due to a flooding or severe weather initiating event, which then required a Phase 3 analysis. The Phase 3 analysis calculated a change in core damage frequency of 3.04E-8/yr, which represented very low safety significance. Inspection Report# : 2009002 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation for Standby Service Water Cooling Tower Drift Eliminators The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, involving a failure to perform an adequate operability evaluation. The inspectors identified non-conservatisms in the evaluation with regards to standby service water cooling tower drift rate, a failure to consider external events design basis impacts, and a failure to properly classify the condition as a substantially degraded, non-conforming condition, because it was subsequently determined that the deficiency could increase drift losses by a factor of ten. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-01222. This finding is more than minor because the failure to perform adequate operability evaluations, if left uncorrected, could become a more significant safety concern because the loss rates could become worse over time. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions. Inspection Report# : 2009002 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: FIN Finding Inadequate Fire Drill Critique The inspectors identified a finding for fire brigade performance deficiencies that were not identified by the licensee during a fire drill critique. The inspectors identified several deficiencies during the drill including issues relating to command and control, fire fighting strategy and use of fire fighting equipment. The inspectors provided feedback to plant personnel on the identified performance issues and the inadequate drill evaluation. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 06522. This finding was more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone objective and adversely affected the cornerstone objective to ensure the

availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, was used to analyze the finding since the inadequate critique had an adverse effect on fire brigade effectiveness, in relation to defense-in-depth strategies. Manual Chapter 0609, Appendix F states that findings associated with the onsite manual fire brigade are excluded. Therefore, in accordance with Manual Chapter 0609, the safety significance was determined by regional management review. Regional management concluded that the finding was of very low safety significance because it reflected fire brigade performance during a training drill, rather than during an actual fire. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to have a low enough threshold in identifying performance issues associated with a plant fire drill [P.1(a)]. (Section 1R05) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of the Engineering Safety Features Electrical Switchgear and Battery Room Ventilation System The inspectors identified a noncited violation of 10 CFR 50.65(a)(2) for the failure to adequately monitor the performance of the engineering safety features electrical switchgear and battery room ventilation system. The inspectors identified a condition report from March 2004 that had not been screened and evaluated in the maintenance rule database as a maintenance preventable functional failure. The condition report identified a room cooler that had tripped due to excessive current on the fan motor because an incorrectly sized sheave was installed during previous maintenance. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 02219. The inspectors determined that this finding was more than minor since the engineering safety features electrical switchgear and battery room ventilation system was not placed in (a)(1) monitoring status in a timely manner. In addition, the finding was more than minor since violations of 10 CFR 50.65(a)(2) necessarily involve degraded system performance, which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the maintenance rule aspect of the finding did not lead to an actual loss of safety function of the system or cause a component to be inoperable, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. (Section 1R12) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of Inadequate Operability Evaluations The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving two examples of a failure to follow procedures which resulted in inadequate operability evaluations. The first example involved an inadequate evaluation of foreign material in the condensate storage tank. The evaluation relied on an assumption that the high-pressure core spray and reactor core isolation cooling pumps would not be damaged by metal debris entrained in the pumps suction. The second example involved an inadequate evaluation of the structural integrity of the standby service water cooling towers that only considered the loss of structural support from a single beam. The licensee entered these issues into the corrective action program as Condition Reports CR GGN 2008 05685 and CR GGN 2008 06044. This finding is more than minor because the failure to perform adequate operability evaluations, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions [H.1(b)]. (Section 1R15

Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Standby Service Water Corrosion The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving a failure to take corrective actions to prevent recurrence of severe corrosion in piping hangers, piping supports, and piping in the standby service water basin cooling towers. Significant corrosion of the standby service water supports in October 2008 had been previously identified by plant personnel during a ten-year in-service inspection on October 3, 1993. At that time, plant personnel determined this to be a significant degraded condition of a safety related system, requiring replacement of the piping and associated supports. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-05434. This finding was more than minor because the corrosion represented a degrading condition that if left uncorrected could become more significant safety concern. The finding was also more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not represent an actual loss of safety function, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Following Identification of Degrading Standby Service Water Supports The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, involving the failure to take timely corrective actions for corrosion on distribution beam structural support posts in the standby service water basin cooling towers. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-05434. The finding was more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance since it did not represent an actual loss of safety function of the standby service water cooling towers, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because licensee personnel failed to identify issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Nov 06, 2008 Identified By: NRC Item Type: NCV NonCited Violation B.5.b. Phase 2 and 3 Mitigating Strategy This finding, affecting the Mitigating Systems Cornerstone, is related to mitigative measures developed to cope with losses of large areas of the plant; in response to Section B.5.b. of the February 25, 2002, Interim Compensatory Measures (ICM) Order (EA-02-026) and related NRC guidance. This finding has been designated as "Official Use

Only - Security-Related Information;" therefore, the details of this finding are being withheld from public disclosure. This finding has no cross-cutting aspect. See inspection report 2008-007 for more details. Inspection Report# : 2008007 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of Four Maintenance Rule Systems. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(2) involving the failure to adequately monitor the performance of four Maintenance Rule systems. Several discrepancies in the Maintenance Rule Program were discovered by the inspectors, including unevaluated condition monitoring failures in the neutron monitoring system and an unevaluated functional failure in the standby gas treatment system. Plant personnel implemented additional corrective actions to fully investigate the potential extent of this condition and the apparent weakness in the condition report screening process used for the Maintenance Rule program. As a result, the Maintenance Rule expert panel classified four systems as needing increased monitoring and goal setting, moving these systems from an a(2) to an a(1) status. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-02219. This finding is more than minor since it was similar to Inspection Manual Chapter 0612, Appendix E, Example 7.b in that the problem involved degraded equipment performance. This finding was characterized under the significance determination process as having very low safety significance because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because licensee personnel failed to use proper self-checking and peer-checking to identify repetitive maintenance rule functional failures and also failed to properly document condition report screening activities [H.4(a)] (Section 1R12). Inspection Report# : 2008004 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment during Adverse Weather Conditions. The inspectors identified a Green noncited violation of 10 CFR 50.65 (a)(4), involving the failure to perform risk assessments following multiple declared tornado watches affecting Grand Gulf Nuclear Station during the landfall of Hurricane Gustav. On the morning of September 3, 2008, the inspectors noted that the licensee had not evaluated the increased risk from a declared tornado watch for the Claiborne County area. The inspectors brought this to the attention of plant personnel and a risk assessment was performed and plant risk was changed from a Green to a Yellow risk condition. The inspectors then reviewed the tornado watches declared by the National Weather Service that affected Claiborne County during the landfall of Hurricane Gustav, and noted that six separate tornado watches had been declared over the previous three days. A review of the control room logs showed no documentation of changes in plant risk condition. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-04397. This finding is more than minor because the risk assessments failed to consider unusual external conditions that were present or imminent. Using Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, Assessment of Risk Deficit and consulting with the regional senior risk analyst, the inspectors determined the finding of very low safety significance due to a calculated incremental core damage probability deficit of 4.38E-08. This finding has a crosscutting aspect in the area of human performance associated with work practices in that plant personnel failed to follow the risk management procedure [H.4(b)] (Section 1R13). Inspection Report# : 2008004 (pdf) Significance: Sep 21, 2008 Identified By: NRC Item Type: VIO Violation

Failure to Perform an Adequate Inspection of PMP Door Seals Protecting Safety Related Equipment. The inspectors identified a Green cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. The licensee had previously received a noncited violation for inadequate inspections of probable maximum precipitation door seals in NRC Inspection Report 05000416/2008002. On July 9, 2008, the inspectors found the entrance door to the Train B standby service water pump house not meeting the standards of the maintenance procedure because the door seals failed to make contact with the door. The extent of condition review found seven additional door seals degraded, including the doors to the diesel generator building and control building. The door seal on the Train B standby service water pump house identified by the inspectors on July 9, 2008, had not been identified by plant personnel during an extent of condition review on February 29, 2008. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-03216. The finding is more than minor since it affects the protection against external factors attribute of mitigating system cornerstone. The door seals also represent a degrading condition that if left uncorrected could affect the availability, reliability, and capability of mitigating systems required to respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multi-trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis using many bounding and conservative assumptions. The result was a delta-CDF of 3.3E 7/yr and a delta-LERF of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution in that the licensee failed to take adequate corrective actions to ensure degraded probable maximum precipitation door seals were properly evaluated and repaired in a timely manner [P.1(d)] (Section 4OA2). Inspection Report# : 2008004 (pdf) Inspection Report# : 2009002 (pdf) Barrier Integrity Significance: Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Fully Close a LPCS Manual Valve Resulted in Leakage of Water into the Condensate and Refueling Water Storage System The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1(a) involving a failure to implement the low pressure core spray system operating instruction correctly. On April 20, 2009, the site was performing a low pressure core spray quarterly surveillance. During the test, the suppression pool level lowered approximately 0.8 inches, which equates to approximately 3600 gallons of water. Plant personnel investigated these anomalies and determined that the low pressure core spray pump had pressurized the condensate and refuelling water storage system due to a partially opened manual fill valve. This valve is a chain-fall operated valve and was approximately five turns open. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2009-02073. The finding was more than minor because it was associated with configuration control attribute of the reactor safety barrier integrity cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from the radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it only represents a degradation of the radiological barrier function provided for the auxiliary building. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources in that the operators did not have specific training in chain-fall type valve operation [H.2(b)] (Section 1R22). Inspection Report# : 2009003 (pdf)

Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Enter a Limiting Condition for Operation for Primary Containment Isolation Valves The inspectors identified a Green noncited violation of Grand Gulf Nuclear Station Technical Specifications 3.6.1.3, for failure to enter a limiting condition for operation action statement for primary containment isolation valves. As a result, the limiting condition for operation action statement time was exceeded. The inspectors identified that surveillance test data for the residual heat removal Train A minimum flow valve was missing. The inspectors discovered that operations staff failed to properly review the work order for the valve work, and they had made an assumption the work order had been canceled. The licensee reviewed the identified issue for extent of condition and identified that in addition to a missed postmaintenance stroke test, they had also failed to enter the limiting condition for operation for two containment isolation valves. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-01069. This finding was more than minor since it affects the configuration control attribute of barrier integrity cornerstone, in that failing to properly test containment isolation valves could affect the assurance that physical design barriers that protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the containment system. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices, in that the operations shift supervisor and maintenance coordinator failed to perform proper self- and peer-checking and proper documentation of the completed work activity. Inspection Report# : 2009002 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Leaking Reactor Water Cleanup System Primary Containment Isolation Valves The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, involving the failure to correct leaking reactor water cleanup system primary containment isolation valves. During refuelling Outage 16, plant personnel were performing local leak rate testing of reactor water cleanup backwash containment penetration. Testing determined that these primary containment isolation valves exceeded the allowable leakage rate by greater than 10 times the leakage limits. The inspectors determined that for four consecutive operating cycles, the site had failed to take corrective actions to correct the excessive leakage through these valves. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 05139. The finding was more than minor because it was associated with systems, structures, and components and the reactor coolant system barrier performance attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the containment system. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee failed to take actions to correct a long-standing equipment issue associated with excessive leakage from primary containment isolation valves [H.2 (a)]. (Section 1R20) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Monitor Plant Parameters to Control Reactor Coolant System Cooldown Rate The inspectors identified a Green finding involving the failure to demonstrate proper monitoring of plant parameters to control reactor coolant system cooldown rate to within expected management standards. The plant experienced a

reactor scram from approximately 15 percent power during plant start-up from a refuelling outage due to a total loss of feedwater. Reactor pressure decreased at a faster rate than expected due to low decay heat levels and the injection of relatively cold condensate storage tank water to reactor vessel. The control room supervisor did not give a pressure band after pressure decreased below the low end of the emergency operating procedure band of 800 psig or assign a licensed operator to monitor reactor pressure during the event. The inspectors identified to the operators that the plant was approaching the procedural limit for cooldown rate; operators then closed the inboard main steam isolation valves to prevent exceeding the cooldown rate. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 06201. The finding is more than minor since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not represent an actual degradation of the radiological barrier function of the reactor coolant system barrier. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because control room supervision failed to maintain proper oversight to ensure reactor coolant cooldown rate was maintained within procedural limits [H.4(c)]. Inspection Report# : 2008005 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : August 31, 2009

Grand Gulf 1 3Q/2009 Plant Inspection Findings Initiating Events Significance: Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Respond to Control Room Alarms in Accordance with Plant Procedures A self-revealing noncited violation of Technical Specification 5.4.1(a) was reviewed involving a failure to follow the fire alarm response procedure during a fire in the reactor feedwater pump area. The operators failed to investigate the source of a smoke alarm for an hour, allowing a fire to develop beyond the incipient stage before it was discovered. On November 17, 2008, a fire ignited in oil-soaked insulation on the reactor feedwater Pump B. After two weeks of plant operation following a refueling outage, during the November 17 shift turnover meeting, a fire alarm was received in the control room and was acknowledged by an operator. No notifications were made to the shift manager, and no operator or fire brigade member was dispatched. One hour after shift turnover, during normal operator rounds the turbine building operator discovered the fire in the reactor feedwater pump room. The fire brigade was dispatched to extinguish the fire. The licensee entered this condition in the corrective action program as condition report CR-GGN-2008-06584. The finding was more than minor because it was associated with the initiating events cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The finding was assessed by performing a bounding analysis using Appendix M of Inspection Manual Chapter 0609. The bounding analysis indicated that the change in core damage frequency would be 4.24 x 10-7 over a 1-year assessment period, indicating that this finding was of very low safety significance. This finding has a crosscutting aspect in the area of human performance with a work practices component for failure to use proper self checking techniques commensurate with the risk of the assigned task to ensure the work is performed safely because operators failed to use self checking techniques when acknowledging the reactor feedwater pump fire alarm [H.4(a)] (Section 4OA3). Inspection Report# : 2009003 (pdf) Significance: Feb 12, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Procedure Requirements for Preventive Maintenance Strategy Development A Green self revealing finding was identified for the failure to implement maintenance procedure requirements. Specifically, in June 2007, an incorrect preventive maintenance template was applied to the main transformer auxiliary power transfer switch resulting in a less than optimal preventive maintenance strategy. This was subsequently determined to be a contributing cause to the January 12th reactor scram. This issue is entered in the corrective action program as condition Report 2008 0174. The performance deficiency associated with this finding is the failure of maintenance and engineering personnel to implement the requirements of Procedure EN-DC-335, PM Basis Template, Section 5.2, PM Basis Template Development. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with work practices, in that the supervisory and management oversight of work activities were not employed such that nuclear safety was supported [H.4.(c)] (Section 4OA4).

Inspection Report# : 2009007 (pdf) Significance: Feb 12, 2009 Identified By: NRC Item Type: FIN Finding Failure to Implement Procurement Engineering Procedure Requirements A Green self revealing finding was identified for the failure of engineering and maintenance personnel to implement procurement engineering procedure requirements. Specifically, in January, 2007 a procurement engineering evaluation determined that a difference in part numbers provided by a vendor was an administrative part number change. Consequently, a current transformer with a slightly different form, fit, and operating characteristic was installed in the generator/unit differential trip circuitry. This combined with other unknown circuit deficiencies and grid reactive load anomalies, resulted in a generator trip and reactor scram on March 21, 2008. The finding is entered in the corrective action program as Condition Report 2008-01476. The performance deficiency associated with this finding is the failure of procurement engineering personnel to implement the requirements of Procedure EN-DC-313, Procurement Engineering Process, Section 5.6, Administrative Part Number Changes, resulting in a less than optimal replacement part for a current transformer in the Unit/Generator differential trip circuitry. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with decision making in that procurement engineering did not use conservative assumptions and adopt a requirement to demonstrate a proposed action is safe to proceed rather than to demonstrate that an action is unsafe to disprove the action [H.1.(b)] (Section 4OA4). Inspection Report# : 2009007 (pdf) Significance: Feb 12, 2009 Identified By: NRC Item Type: FIN Finding Failure to Implement Preventive Maintenance Procedure Requirements A Green self revealing finding was identified for the failure of to implement maintenance procedure requirements. Specifically, between 2002 and 2008, neither the preventive maintenance optimization program, nor the turbine 10-year plan prescribed a preventive maintenance strategy for the thyristor voltage regulator control portion of the main generator voltage regulating system. Consequently, on October 26, 2008, an under-excitation condition existed in the main generator following transfer from automatic to manual voltage regulator control, resulting in a generator and turbine trip and a reactor scram. The finding is entered in the corrective action program as Condition Report 2008-6241. The performance deficiency associated with this finding is the failure of maintenance and engineering personnel to implement the requirements of Procedure EN-DC-324, Preventive Maintenance Programs, Section 5.2, Process Overview, and Procedure EN-DC-335, PM Basis Template, Section 5.2, PM Basis Template Development. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with decision making, in that a systematic process was not employed for risk significant decision making and that roles and authority for decision making was not formally defined [H.1.(a)] (Section 4OA4). Inspection Report# : 2009007 (pdf)

Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Trip of a Reactor Recirculation Pump During Pump Up-shift to Fast Speed Due to Ineffective Corrective Actions The inspectors reviewed a self-revealing Green finding involving a recirculation pump trip during pump up-shift to fast speed due to ineffective corrective actions. The plant had recently replaced the recirculation motor on Pump A during the refuelling outage and during investigation determined that the instantaneous over-current trip for the breaker had drifted low. The inspectors performed a review of condition reports and determined that reactor recirculation Pump B had tripped following motor replacement for the same reason in September 2007. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-06269. The finding was more than minor because it was associated with the initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not contribute to loss of function of mitigating equipment. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program in that the licensee failed to perform a thorough evaluation of a problem that resulted in a plant transient such that the resolution properly addressed the cause and extent of condition [P.1(c)]. (Section 1R20) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: Self-Revealing Item Type: FIN Finding Automatic Reactor Scram Caused by an Operator Inadvertently Closing Steam Supply Valves to the Reactor Feed Pump Turbine The inspectors reviewed a self-revealing Green finding involving an automatic reactor scram caused by an operator inadvertently closing steam supply valves to the reactor feed pump turbine. Site personnel investigating the scram determined that an operator had incorrectly performed actions for the reactor feed Pump B turbine on the reactor feed Pump A turbine control switches at a local panel. The operator inadvertently closed the steam supply valves to the reactor feed Pump A turbine resulting in a total loss of feedwater flow and low reactor water level scram. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-06195. The finding was more than minor because it was associated with the initiating events cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that an evaluation was required by the regional senior reactor analyst, because the finding impacted both the initiating event and mitigating systems cornerstone. The senior reactor analyst performed a Phase 3 analysis and determined the issue was very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the operator failed to use proper self-checking techniques while performing actions to place feed Pump B in the standby lineup [H.4(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Mitigating Systems Significance: Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of a Maintenance Rule Scoped System Green. The inspectors identified a Green noncited violation of 10 CFR Part 50.65(a)(2) involving the failure to

adequately monitor the performance of a maintenance rule scoped system. The licensees maintenance rule program required evaluation of the area radiation monitoring system for classification as a maintenance rule (a)(1) system after three failures within eighteen months. The licensee had identified two functional failures of the residual heat removal heat exchanger A hatch radiation monitor in June and July 2008. The inspectors identified three other instances of functional failures on components that were used in plant emergency operating procedures and emergency preparedness procedures. These failures were not included in the licensees maintenance rule database. A total of five functional failures occurred in system components before the licensee considered evaluation of area radiation monitoring as a maintenance rule (a)(1) system in September 2009. The licensee entered this condition in the corrective action program as condition reports CR-GGN-2009-04853 and CR-GGN-2009-04857. The finding was more than minor because it was similar to Inspection Manual Chapter 0612, Appendix E, Example 7.d, in that equipment performance problems were such that effective control of performance or condition through appropriate preventive Maintenance Under (a)(2) could not be demonstrated. In addition, it affected the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was characterized under the significance determination process as having very low safety significance because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue does not reflect current licensee performance. (Section 1R12) Inspection Report# : 2009004 (pdf) Significance: Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Operator Response Times to Fires Green. The inspectors identified a Green non-cited violation of Technical Specification 5.4.1(a), for failure to ensure that operators can respond in timely manner to safe shutdown panels in the auxiliary building with a fire in the main control room. The inspectors reviewed a condition report associated with response times of operators to a fire in the protected area with Mississippi river at flood stage. The inspectors questioned the adequacy of response times for fire brigade members and the safe shutdown operator in the event of fire in the control room with the designated operators being outside the protected area. The licensee determined a time critical task would not have been completed due to the safe shutdown operator being outside the protected area. The licensee entered this condition in the corrective action program as condition report CR-GGN-2009-01416. The inspectors determined this finding to be more than minor since it affected the external events 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. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, it was determined that the finding screened as potentially risk significant due to external events and required the regional senior reactor analyst to perform a Phase 3 evaluation. The senior reactor analyst determined the likelihood that control room abandonment occurs while the safe shutdown operator is out of the protected area is 9.78E-8. The change in core damage frequency is lower than this value and small enough that large early release frequency is not required to be considered. Therefore the issue is (Green) of very low safety significance. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with corrective action program in that the licensee failed to perform an appropriate extent of condition when implementing corrective action associated with fire brigade response issue in 2008 [P.1(c)]. (Section 4OA2) Inspection Report# : 2009004 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Design Changes to Protect the Standby Service Water Slab The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III involving the failure to incorporate design changes required to limit dynamic loads on the standby service water basin slab. In 1997, the plant

experienced damage to the standby service water basin slabs resulting from unanalyzed dynamic loads. During a standby service water system inspection on April 18, 2009, inspectors observed several different tire tracks on the seismically-designed concrete slab that covers and is integral to the safety-related standby service water basin. The inspectors also noted small placards attached to the basin slabs which prohibited moving vehicles on the slabs, and other signs requiring protective mats under any items placed on the slabs. Plant personnel evaluated the loading of the vehicle and determined that the load limits on the basin slab had not been exceeded. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2009 002087. The inspectors determined this finding affected the design control 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. In addition, the finding was more than minor because the failure to prevent dynamic loads on the standby service water basin slabs, if left uncorrected, could become more significant safety concern. Using the Manual Chapter of 0609, Significance Determination Process, Phase 1 Worksheet, this finding was determined to have very low safety significance, because it did not represent an actual loss of a safety function of the standby service water system. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue occurred several years ago and does not reflect current licensee performance (Section 1R04). Inspection Report# : 2009003 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation for Debris Left in the Condensate Storage Tank The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving a failure to follow procedures which resulted in an inadequate operability evaluation. During the week of May 18, 2009, the site conducted debris removal in the condensate storage tank. This debris removal was necessary because of a failure to remove all debris in the condensate storage tank during their spring 2007 cleanup project. The licensee performed an operability evaluation for objects left in the condensate storage tank which stated that the high pressure core spray system and reactor core isolation cooling would remain operable for all postulated events. Upon review by the inspectors, the operability evaluation did not address several issues including objects left in the condensate storage tank and condensate system return flow to the condensate storage tank following a plant shutdown/scram. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2009-02815 and CR-GGN-2009-02837. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to thoroughly identify all materials left in the condensate storage tank during their original operability determination [P.1(a)] (Section 1R15). Inspection Report# : 2009003 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: FIN Finding Failure to Perform a Timely Operability Evaluation Following the Discovery of a SSW Fan Failure Mechanism The inspectors identified a Green finding involving the failure to perform an operability determination after a new failure mechanism was discovered for standby service water Fan B. The inspectors were performing a follow up review of a condition report that detailed a trip of Division 1 standby service water Fan B. The fan had tripped on start up from the control room on December 31, 2007. The licensee had initially determined the trip was due to a faulty solid state trip device. Subsequent testing in failed to identify a problem with the trip device, and the apparent cause of the fan trip was attributed to reverse rotation of the fan. Operations personnel were not informed of this new information as required by the corrective action program procedure. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2009-01726.

This finding is more than minor because it was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that the finding had very low safety significance (Green) since it did not represent an actual loss of a safety function of the standby service water cooling towers, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of human performance associated with work practices in that licensee personnel failed to apply procedural requirements to write a new condition report when new information was acquired related to the trip of the Division 1 standby service water Fan B [H.4(b)] (Section 4OA2). Inspection Report# : 2009003 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of Division II Standby Diesel Generator Jacket Water Heat Exchanger The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," for failure to comply with the licensees Generic Letter 89 13 program, which specifically states that cleaning of heat exchangers covered by this program is prohibited prior to performing an as-found thermal performance test. Specifically, in early 2006, the Division II Standby Diesel Generator (i.e. Emergency Diesel Generator) jacket water cooling heat exchanger was cleaned just prior to performing a five year thermal performance test. The licensee has entered this into their corrective action program as CR-GGN-2009-00904. This finding is more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of safety systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The inspectors reviewed the finding for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Bias in Instrumentation Used for Standby Service Water Leak Detection The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to establish adequate measures for the selection and review for suitability of equipment and processes that are essential to the safety-related functions of structures, systems and components. Specifically, the licensee failed to properly design for pulsation effects on flow rate instrumentation used for leak detection in the Standby Service Water system. This instrumentation is needed to meet licensee commitment 10 CFR Part 50, Appendix A, General Design Criterion 13, "Instrumentation and Control," to monitor trends in the ultimate heat sink basin inventory with the system in operation. The system was designed to detect a leakage rate of 1250 gallons per minute, and alarm in the control room at this leak rate, but due to design inadequacies in the instrumentation, the leak rate would have to exceed 3350 gallons per minute before activating the alarm. The licensee has entered this into their corrective action program as CR GGN 2009-00054. This finding was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1 3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter

0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The finding was reviewed for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Motor-Operated Valve Calculations Used Non-Conservative Inputs and Methodologies The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" for failing to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee used non conservative inputs or methodologies in calculating terminal voltages to safety related motor-operated valve motors that would be required to operate for mitigation of design bases events. The licensees electrical calculations used non conservative 50 percent locked rotor currents and neglected thermal overload resistance to determine the terminal voltages to safety related motor-operated valves which would predict higher terminal voltages than would actually exist. The calculated terminal voltages were direct design inputs into the applicable motor-operated valves mechanical thrust and torque calculations. The licensee has entered this issue into their corrective action program as CR GGN 2009 00985. This finding was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1 3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. This finding has a cross cutting aspect in the area of Problem Identification and Resolution, in that self assessments are of sufficient depth, are comprehensive, are appropriately objective, and are self critical. The licensee had conducted a Component Design Bases Assessment, LO GLO 2008 00044 in August 2008, and failed to adequately assess an identical finding identified at River Bend Station during their 2008 Component Design Bases Inspection. The licensee had determined that this issue was not applicable at Grand Gulf Nuclear Station. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions fo rReplacement of Safety-Related Batteries The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for failure to identify and correct a condition adverse to quality related to the seismic qualification of the Division III High Pressure Core Spray safety-related battery. Specifically, the licensee failed to identify an incorrectly installed end bracket after replacement of the Division III safety-related battery in 2002 using procedures, work instructions, and drawings that were supposed to have been corrected after this same issue was identified during a 1997 battery replacement activity. The licensee has entered this into their corrective action program as CR-GGN-2009-00830. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events for ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was confirmed to not result in a loss of operability or functionality. The finding was reviewed for cross-cutting aspects and none were identified.

Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of a Failure to Meet 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," with two examples. Specifically, the team identified that the licensee failed to develop and implement adequate testing programs for Class 1E molded-case circuit breakers, and for the voltage and frequency response of the standby diesel generators that met design or vendor requirements and recommendations. In response, the licensee entered these examples in the corrective action program as CR GGN 2009 01024, and CR GGN-2009-01057. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events for ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, each example was determined to be of very low safety significance (Green) because they did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding has a cross cutting aspect in the area of Problem Identification and Resolution, in that self assessments are of sufficient depth, are comprehensive, are appropriately objective, and are self critical. The licensee had conducted a Component Design Bases Assessment, LO-GLO-2008-00044 in August 2008, and failed to adequately assess an identical finding identified at River Bend Station during their 2008 Component Design Bases Inspection. The licensee had determined that this issue was not applicable at Grand Gulf Nuclear Station. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control for Standby Service Water Pump Cables and Electrical Vaults The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to adequately demonstrate operability for the 4160 volt Standby Service Water Pump kerite cables through adequate testing and analysis in a continuously submerged environment. Furthermore, the environment for these continuously submerged cables exists because each of the two vaults that contain these cables (MH 20 and MH 21) has a design flaw, in that several other vaults gravity drain to them and the design of these vaults did not include a sump pump or other means for water to be removed or drained from them. The licensee has entered this into their corrective action program as CR-GGN-2009-01028. This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control of ensuring the availability, reliability, and capability of safety systems, and closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, because there was reasonable doubt on the continued operability of the Standby Service Water system. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution in that the licensee failed to implement Operating Experience directly communicated with a Generic Letter through changes to station processes, procedures, and equipment. Inspection Report# : 2009006 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedures to Maintain Drains on Safety Related Buildings The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, involving the failure

to properly clean and inspect the rooftop and associated water drainage systems of the safety-related diesel generator building. The inspectors identified loose, flexible roofing material that could have covered roof drains and result in loss of functionality for all of the standby diesel generators during a design basis heavy rainfall event. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-00429. This finding is more than minor since it affects the protection against external events attribute of mitigating system cornerstone. The roofing material and debris represented a degrading condition that if left uncorrected could have affected the availability, reliability, and capability of the standby diesel generators to respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding screened as potentially risk significant due to a flooding or severe weather initiating event, which then required a Phase 3 analysis. The Phase 3 analysis calculated a change in core damage frequency of 3.04E-8/yr, which represented very low safety significance. Inspection Report# : 2009002 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation for Standby Service Water Cooling Tower Drift Eliminators The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, involving a failure to perform an adequate operability evaluation. The inspectors identified non-conservatisms in the evaluation with regards to standby service water cooling tower drift rate, a failure to consider external events design basis impacts, and a failure to properly classify the condition as a substantially degraded, non-conforming condition, because it was subsequently determined that the deficiency could increase drift losses by a factor of ten. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-01222. This finding is more than minor because the failure to perform adequate operability evaluations, if left uncorrected, could become a more significant safety concern because the loss rates could become worse over time. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions. Inspection Report# : 2009002 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: FIN Finding Inadequate Fire Drill Critique The inspectors identified a finding for fire brigade performance deficiencies that were not identified by the licensee during a fire drill critique. The inspectors identified several deficiencies during the drill including issues relating to command and control, fire fighting strategy and use of fire fighting equipment. The inspectors provided feedback to plant personnel on the identified performance issues and the inadequate drill evaluation. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 06522. This finding was more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone objective and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, was used to analyze the finding since the inadequate critique had an adverse effect on fire brigade effectiveness, in relation to defense-in-depth strategies. Manual Chapter 0609, Appendix F states that findings associated with the onsite manual fire brigade are excluded. Therefore, in accordance with Manual Chapter 0609, the safety significance was determined by regional management review. Regional management concluded that the finding was of very low safety significance because it reflected fire brigade performance during a training drill, rather than during an actual fire. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to have a low enough threshold in identifying performance issues associated with a plant fire drill [P.1(a)]. (Section 1R05)

Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of the Engineering Safety Features Electrical Switchgear and Battery Room Ventilation System The inspectors identified a noncited violation of 10 CFR 50.65(a)(2) for the failure to adequately monitor the performance of the engineering safety features electrical switchgear and battery room ventilation system. The inspectors identified a condition report from March 2004 that had not been screened and evaluated in the maintenance rule database as a maintenance preventable functional failure. The condition report identified a room cooler that had tripped due to excessive current on the fan motor because an incorrectly sized sheave was installed during previous maintenance. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 02219. The inspectors determined that this finding was more than minor since the engineering safety features electrical switchgear and battery room ventilation system was not placed in (a)(1) monitoring status in a timely manner. In addition, the finding was more than minor since violations of 10 CFR 50.65(a)(2) necessarily involve degraded system performance, which, if left uncorrected, could become a more significant safety concern. This finding has very low safety significance because the maintenance rule aspect of the finding did not lead to an actual loss of safety function of the system or cause a component to be inoperable, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. (Section 1R12) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of Inadequate Operability Evaluations The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving two examples of a failure to follow procedures which resulted in inadequate operability evaluations. The first example involved an inadequate evaluation of foreign material in the condensate storage tank. The evaluation relied on an assumption that the high-pressure core spray and reactor core isolation cooling pumps would not be damaged by metal debris entrained in the pumps suction. The second example involved an inadequate evaluation of the structural integrity of the standby service water cooling towers that only considered the loss of structural support from a single beam. The licensee entered these issues into the corrective action program as Condition Reports CR GGN 2008 05685 and CR GGN 2008 06044. This finding is more than minor because the failure to perform adequate operability evaluations, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions [H.1(b)]. (Section 1R15 Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Standby Service Water Corrosion The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, involving a failure to take corrective actions to prevent recurrence of severe corrosion in piping hangers, piping supports, and piping in the standby service water basin cooling towers. Significant corrosion of the standby service water supports in October

2008 had been previously identified by plant personnel during a ten-year in-service inspection on October 3, 1993. At that time, plant personnel determined this to be a significant degraded condition of a safety related system, requiring replacement of the piping and associated supports. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-05434. This finding was more than minor because the corrosion represented a degrading condition that if left uncorrected could become more significant safety concern. The finding was also more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not represent an actual loss of safety function, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Untimely Corrective Actions Following Identification of Degrading Standby Service Water Supports The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, involving the failure to take timely corrective actions for corrosion on distribution beam structural support posts in the standby service water basin cooling towers. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2008-05434. The finding was more than minor because it was associated with the protection against external factors attribute of the reactor safety mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance since it did not represent an actual loss of safety function of the standby service water cooling towers, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because licensee personnel failed to identify issues completely, accurately, and in a timely manner commensurate with their safety significance [P.1(a)]. (Section 4OA3) Inspection Report# : 2008005 (pdf) Significance: Nov 06, 2008 Identified By: NRC Item Type: NCV NonCited Violation B.5.b. Phase 2 and 3 Mitigating Strategy This finding, affecting the Mitigating Systems Cornerstone, is related to mitigative measures developed to cope with losses of large areas of the plant; in response to Section B.5.b. of the February 25, 2002, Interim Compensatory Measures (ICM) Order (EA-02-026) and related NRC guidance. This finding has been designated as "Official Use Only - Security-Related Information;" therefore, the details of this finding are being withheld from public disclosure. This finding has no cross-cutting aspect. See inspection report 2008-007 for more details. Inspection Report# : 2008007 (pdf) Barrier Integrity Significance: Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation

Failure to Fully Close a LPCS Manual Valve Resulted in Leakage of Water into the Condensate and Refueling Water Storage System The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1(a) involving a failure to implement the low pressure core spray system operating instruction correctly. On April 20, 2009, the site was performing a low pressure core spray quarterly surveillance. During the test, the suppression pool level lowered approximately 0.8 inches, which equates to approximately 3600 gallons of water. Plant personnel investigated these anomalies and determined that the low pressure core spray pump had pressurized the condensate and refuelling water storage system due to a partially opened manual fill valve. This valve is a chain-fall operated valve and was approximately five turns open. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2009-02073. The finding was more than minor because it was associated with configuration control attribute of the reactor safety barrier integrity cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from the radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it only represents a degradation of the radiological barrier function provided for the auxiliary building. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources in that the operators did not have specific training in chain-fall type valve operation [H.2(b)] (Section 1R22). Inspection Report# : 2009003 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Enter a Limiting Condition for Operation for Primary Containment Isolation Valves The inspectors identified a Green noncited violation of Grand Gulf Nuclear Station Technical Specifications 3.6.1.3, for failure to enter a limiting condition for operation action statement for primary containment isolation valves. As a result, the limiting condition for operation action statement time was exceeded. The inspectors identified that surveillance test data for the residual heat removal Train A minimum flow valve was missing. The inspectors discovered that operations staff failed to properly review the work order for the valve work, and they had made an assumption the work order had been canceled. The licensee reviewed the identified issue for extent of condition and identified that in addition to a missed postmaintenance stroke test, they had also failed to enter the limiting condition for operation for two containment isolation valves. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-01069. This finding was more than minor since it affects the configuration control attribute of barrier integrity cornerstone, in that failing to properly test containment isolation valves could affect the assurance that physical design barriers that protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the containment system. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices, in that the operations shift supervisor and maintenance coordinator failed to perform proper self- and peer-checking and proper documentation of the completed work activity. Inspection Report# : 2009002 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Leaking Reactor Water Cleanup System Primary Containment Isolation Valves The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion XVI, involving the failure to correct leaking reactor water cleanup system primary containment isolation valves. During refuelling Outage 16, plant personnel were performing local leak rate testing of reactor water cleanup backwash containment penetration. Testing determined that these primary containment isolation valves exceeded the allowable leakage rate by greater than 10 times the leakage limits. The inspectors determined that for four consecutive operating cycles, the

site had failed to take corrective actions to correct the excessive leakage through these valves. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 05139. The finding was more than minor because it was associated with systems, structures, and components and the reactor coolant system barrier performance attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the containment system. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee failed to take actions to correct a long-standing equipment issue associated with excessive leakage from primary containment isolation valves [H.2 (a)]. (Section 1R20) Inspection Report# : 2008005 (pdf) Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Monitor Plant Parameters to Control Reactor Coolant System Cooldown Rate The inspectors identified a Green finding involving the failure to demonstrate proper monitoring of plant parameters to control reactor coolant system cooldown rate to within expected management standards. The plant experienced a reactor scram from approximately 15 percent power during plant start-up from a refuelling outage due to a total loss of feedwater. Reactor pressure decreased at a faster rate than expected due to low decay heat levels and the injection of relatively cold condensate storage tank water to reactor vessel. The control room supervisor did not give a pressure band after pressure decreased below the low end of the emergency operating procedure band of 800 psig or assign a licensed operator to monitor reactor pressure during the event. The inspectors identified to the operators that the plant was approaching the procedural limit for cooldown rate; operators then closed the inboard main steam isolation valves to prevent exceeding the cooldown rate. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2008 06201. The finding is more than minor since it affects the human performance attribute of the barrier integrity cornerstone and affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, inspectors determined that the finding has very low safety significance (Green) since it did not represent an actual degradation of the radiological barrier function of the reactor coolant system barrier. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because control room supervision failed to maintain proper oversight to ensure reactor coolant cooldown rate was maintained within procedural limits [H.4(c)]. Inspection Report# : 2008005 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : December 10, 2009

Grand Gulf 1 4Q/2009 Plant Inspection Findings Initiating Events Significance: Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Respond to Control Room Alarms in Accordance with Plant Procedures A self-revealing noncited violation of Technical Specification 5.4.1(a) was reviewed involving a failure to follow the fire alarm response procedure during a fire in the reactor feedwater pump area. The operators failed to investigate the source of a smoke alarm for an hour, allowing a fire to develop beyond the incipient stage before it was discovered. On November 17, 2008, a fire ignited in oil-soaked insulation on the reactor feedwater Pump B. After two weeks of plant operation following a refueling outage, during the November 17 shift turnover meeting, a fire alarm was received in the control room and was acknowledged by an operator. No notifications were made to the shift manager, and no operator or fire brigade member was dispatched. One hour after shift turnover, during normal operator rounds the turbine building operator discovered the fire in the reactor feedwater pump room. The fire brigade was dispatched to extinguish the fire. The licensee entered this condition in the corrective action program as condition report CR-GGN-2008-06584. The finding was more than minor because it was associated with the initiating events cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The finding was assessed by performing a bounding analysis using Appendix M of Inspection Manual Chapter 0609. The bounding analysis indicated that the change in core damage frequency would be 4.24 x 10-7 over a 1-year assessment period, indicating that this finding was of very low safety significance. This finding has a crosscutting aspect in the area of human performance with a work practices component for failure to use proper self checking techniques commensurate with the risk of the assigned task to ensure the work is performed safely because operators failed to use self checking techniques when acknowledging the reactor feedwater pump fire alarm [H.4(a)] (Section 4OA3). Inspection Report# : 2009003 (pdf) Significance: Feb 12, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Procedure Requirements for Preventive Maintenance Strategy Development A Green self revealing finding was identified for the failure to implement maintenance procedure requirements. Specifically, in June 2007, an incorrect preventive maintenance template was applied to the main transformer auxiliary power transfer switch resulting in a less than optimal preventive maintenance strategy. This was subsequently determined to be a contributing cause to the January 12th reactor scram. This issue is entered in the corrective action program as condition Report 2008 0174. The performance deficiency associated with this finding is the failure of maintenance and engineering personnel to implement the requirements of Procedure EN-DC-335, PM Basis Template, Section 5.2, PM Basis Template Development. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with work practices, in that the supervisory and management oversight of work activities were not employed such that nuclear safety was supported [H.4.(c)] (Section 4OA4).

Inspection Report# : 2009007 (pdf) Significance: Feb 12, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Procurement Engineering Procedure Requirements A Green self revealing finding was identified for the failure of engineering and maintenance personnel to implement procurement engineering procedure requirements. Specifically, in January, 2007 a procurement engineering evaluation determined that a difference in part numbers provided by a vendor was an administrative part number change. Consequently, a current transformer with a slightly different form, fit, and operating characteristic was installed in the generator/unit differential trip circuitry. This combined with other unknown circuit deficiencies and grid reactive load anomalies, resulted in a generator trip and reactor scram on March 21, 2008. The finding is entered in the corrective action program as Condition Report 2008-01476. The performance deficiency associated with this finding is the failure of procurement engineering personnel to implement the requirements of Procedure EN-DC-313, Procurement Engineering Process, Section 5.6, Administrative Part Number Changes, resulting in a less than optimal replacement part for a current transformer in the Unit/Generator differential trip circuitry. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with decision making in that procurement engineering did not use conservative assumptions and adopt a requirement to demonstrate a proposed action is safe to proceed rather than to demonstrate that an action is unsafe to disprove the action [H.1.(b)] (Section 4OA4). Inspection Report# : 2009007 (pdf) Significance: Feb 12, 2009 Identified By: Self-Revealing Item Type: FIN Finding Failure to Implement Preventive Maintenance Procedure Requirements A Green self revealing finding was identified for the failure of to implement maintenance procedure requirements. Specifically, between 2002 and 2008, neither the preventive maintenance optimization program, nor the turbine 10-year plan prescribed a preventive maintenance strategy for the thyristor voltage regulator control portion of the main generator voltage regulating system. Consequently, on October 26, 2008, an under-excitation condition existed in the main generator following transfer from automatic to manual voltage regulator control, resulting in a generator and turbine trip and a reactor scram. The finding is entered in the corrective action program as Condition Report 2008-6241. The performance deficiency associated with this finding is the failure of maintenance and engineering personnel to implement the requirements of Procedure EN-DC-324, Preventive Maintenance Programs, Section 5.2, Process Overview, and Procedure EN-DC-335, PM Basis Template, Section 5.2, PM Basis Template Development. The finding is more than minor because it is associated with the equipment performance attribute of the initiating events cornerstone and affects the cornerstone objective to limit those events that upset plant stability. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding is determined to have very low safety significance because it did not result in exceeding the technical specification limit for identified reactor coolant system leakage, did not affect mitigation systems, did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and did not increase the likelihood of a fire or internal/external flood. The finding has a cross cutting aspect in the area of human performance associated with decision making, in that a systematic process was not employed for risk significant decision making and that roles and authority for decision making was not formally defined [H.1.(a)] (Section 4OA4). Inspection Report# : 2009007 (pdf)

Mitigating Systems Significance: Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of a Maintenance Rule Scoped System Green. The inspectors identified a Green noncited violation of 10 CFR Part 50.65(a)(2) involving the failure to adequately monitor the performance of a maintenance rule scoped system. The licensees maintenance rule program required evaluation of the area radiation monitoring system for classification as a maintenance rule (a)(1) system after three failures within eighteen months. The licensee had identified two functional failures of the residual heat removal heat exchanger A hatch radiation monitor in June and July 2008. The inspectors identified three other instances of functional failures on components that were used in plant emergency operating procedures and emergency preparedness procedures. These failures were not included in the licensees maintenance rule database. A total of five functional failures occurred in system components before the licensee considered evaluation of area radiation monitoring as a maintenance rule (a)(1) system in September 2009. The licensee entered this condition in the corrective action program as condition reports CR-GGN-2009-04853 and CR-GGN-2009-04857. The finding was more than minor because it was similar to Inspection Manual Chapter 0612, Appendix E, Example 7.d, in that equipment performance problems were such that effective control of performance or condition through appropriate preventive Maintenance Under (a)(2) could not be demonstrated. In addition, it affected the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was characterized under the significance determination process as having very low safety significance because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue does not reflect current licensee performance. (Section 1R12) Inspection Report# : 2009004 (pdf) Significance: Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Operator Response Times to Fires Green. The inspectors identified a Green non-cited violation of Technical Specification 5.4.1(a), for failure to ensure that operators can respond in timely manner to safe shutdown panels in the auxiliary building with a fire in the main control room. The inspectors reviewed a condition report associated with response times of operators to a fire in the protected area with Mississippi river at flood stage. The inspectors questioned the adequacy of response times for fire brigade members and the safe shutdown operator in the event of fire in the control room with the designated operators being outside the protected area. The licensee determined a time critical task would not have been completed due to the safe shutdown operator being outside the protected area. The licensee entered this condition in the corrective action program as condition report CR-GGN-2009-01416. The inspectors determined this finding to be more than minor since it affected the external events 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. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, it was determined that the finding screened as potentially risk significant due to external events and required the regional senior reactor analyst to perform a Phase 3 evaluation. The senior reactor analyst determined the likelihood that control room abandonment occurs while the safe shutdown operator is out of the protected area is 9.78E-8. The change in core damage frequency is lower than this value and small enough that large early release frequency is not required to be considered. Therefore the issue is (Green) of very low safety significance. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with corrective action program in that the licensee failed to perform an appropriate extent of condition when implementing corrective action associated with fire brigade response issue in 2008 [P.1(c)]. (Section 4OA2)

Inspection Report# : 2009004 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Design Changes to Protect the Standby Service Water Slab The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III involving the failure to incorporate design changes required to limit dynamic loads on the standby service water basin slab. In 1997, the plant experienced damage to the standby service water basin slabs resulting from unanalyzed dynamic loads. During a standby service water system inspection on April 18, 2009, inspectors observed several different tire tracks on the seismically-designed concrete slab that covers and is integral to the safety-related standby service water basin. The inspectors also noted small placards attached to the basin slabs which prohibited moving vehicles on the slabs, and other signs requiring protective mats under any items placed on the slabs. Plant personnel evaluated the loading of the vehicle and determined that the load limits on the basin slab had not been exceeded. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2009 002087. The inspectors determined this finding affected the design control 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. In addition, the finding was more than minor because the failure to prevent dynamic loads on the standby service water basin slabs, if left uncorrected, could become more significant safety concern. Using the Manual Chapter of 0609, Significance Determination Process, Phase 1 Worksheet, this finding was determined to have very low safety significance, because it did not represent an actual loss of a safety function of the standby service water system. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue occurred several years ago and does not reflect current licensee performance (Section 1R04). Inspection Report# : 2009003 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation for Debris Left in the Condensate Storage Tank The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving a failure to follow procedures which resulted in an inadequate operability evaluation. During the week of May 18, 2009, the site conducted debris removal in the condensate storage tank. This debris removal was necessary because of a failure to remove all debris in the condensate storage tank during their spring 2007 cleanup project. The licensee performed an operability evaluation for objects left in the condensate storage tank which stated that the high pressure core spray system and reactor core isolation cooling would remain operable for all postulated events. Upon review by the inspectors, the operability evaluation did not address several issues including objects left in the condensate storage tank and condensate system return flow to the condensate storage tank following a plant shutdown/scram. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2009-02815 and CR-GGN-2009-02837. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to thoroughly identify all materials left in the condensate storage tank during their original operability determination [P.1(a)] (Section 1R15). Inspection Report# : 2009003 (pdf) Significance: Jun 23, 2009 Identified By: NRC

Item Type: FIN Finding Failure to Perform a Timely Operability Evaluation Following the Discovery of a SSW Fan Failure Mechanism The inspectors identified a Green finding involving the failure to perform an operability determination after a new failure mechanism was discovered for standby service water Fan B. The inspectors were performing a follow up review of a condition report that detailed a trip of Division 1 standby service water Fan B. The fan had tripped on start up from the control room on December 31, 2007. The licensee had initially determined the trip was due to a faulty solid state trip device. Subsequent testing in failed to identify a problem with the trip device, and the apparent cause of the fan trip was attributed to reverse rotation of the fan. Operations personnel were not informed of this new information as required by the corrective action program procedure. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2009-01726. This finding is more than minor because it was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that the finding had very low safety significance (Green) since it did not represent an actual loss of a safety function of the standby service water cooling towers, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of human performance associated with work practices in that licensee personnel failed to apply procedural requirements to write a new condition report when new information was acquired related to the trip of the Division 1 standby service water Fan B [H.4(b)] (Section 4OA2). Inspection Report# : 2009003 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of Division II Standby Diesel Generator Jacket Water Heat Exchanger The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," for failure to comply with the licensees Generic Letter 89 13 program, which specifically states that cleaning of heat exchangers covered by this program is prohibited prior to performing an as-found thermal performance test. Specifically, in early 2006, the Division II Standby Diesel Generator (i.e. Emergency Diesel Generator) jacket water cooling heat exchanger was cleaned just prior to performing a five year thermal performance test. The licensee has entered this into their corrective action program as CR-GGN-2009-00904. This finding is more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of safety systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The inspectors reviewed the finding for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Bias in Instrumentation Used for Standby Service Water Leak Detection The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to establish adequate measures for the selection and review for suitability of equipment and processes that are essential to the safety-related functions of structures, systems and components. Specifically, the licensee failed to properly design for pulsation effects on flow rate instrumentation used for leak detection in the Standby Service Water system. This instrumentation is needed to meet licensee commitment 10 CFR Part 50, Appendix A, General Design Criterion

13, "Instrumentation and Control," to monitor trends in the ultimate heat sink basin inventory with the system in operation. The system was designed to detect a leakage rate of 1250 gallons per minute, and alarm in the control room at this leak rate, but due to design inadequacies in the instrumentation, the leak rate would have to exceed 3350 gallons per minute before activating the alarm. The licensee has entered this into their corrective action program as CR GGN 2009-00054. This finding was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1 3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The finding was reviewed for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Motor-Operated Valve Calculations Used Non-Conservative Inputs and Methodologies The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" for failing to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee used non conservative inputs or methodologies in calculating terminal voltages to safety related motor-operated valve motors that would be required to operate for mitigation of design bases events. The licensees electrical calculations used non conservative 50 percent locked rotor currents and neglected thermal overload resistance to determine the terminal voltages to safety related motor-operated valves which would predict higher terminal voltages than would actually exist. The calculated terminal voltages were direct design inputs into the applicable motor-operated valves mechanical thrust and torque calculations. The licensee has entered this issue into their corrective action program as CR GGN 2009 00985. This finding was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1 3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. This finding has a cross cutting aspect in the area of Problem Identification and Resolution, in that self assessments are of sufficient depth, are comprehensive, are appropriately objective, and are self critical. The licensee had conducted a Component Design Bases Assessment, LO GLO 2008 00044 in August 2008, and failed to adequately assess an identical finding identified at River Bend Station during their 2008 Component Design Bases Inspection. The licensee had determined that this issue was not applicable at Grand Gulf Nuclear Station. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions fo rReplacement of Safety-Related Batteries The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for failure to identify and correct a condition adverse to quality related to the seismic qualification of the Division III High Pressure Core Spray safety-related battery. Specifically, the licensee failed to identify an incorrectly installed end bracket after replacement of the Division III safety-related battery in 2002 using procedures, work instructions, and drawings that were supposed to have been corrected after this same issue was identified during a 1997 battery

replacement activity. The licensee has entered this into their corrective action program as CR-GGN-2009-00830. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events for ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was confirmed to not result in a loss of operability or functionality. The finding was reviewed for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of a Failure to Meet 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," with two examples. Specifically, the team identified that the licensee failed to develop and implement adequate testing programs for Class 1E molded-case circuit breakers, and for the voltage and frequency response of the standby diesel generators that met design or vendor requirements and recommendations. In response, the licensee entered these examples in the corrective action program as CR GGN 2009 01024, and CR GGN-2009-01057. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events for ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, each example was determined to be of very low safety significance (Green) because they did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding has a cross cutting aspect in the area of Problem Identification and Resolution, in that self assessments are of sufficient depth, are comprehensive, are appropriately objective, and are self critical. The licensee had conducted a Component Design Bases Assessment, LO-GLO-2008-00044 in August 2008, and failed to adequately assess an identical finding identified at River Bend Station during their 2008 Component Design Bases Inspection. The licensee had determined that this issue was not applicable at Grand Gulf Nuclear Station. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control for Standby Service Water Pump Cables and Electrical Vaults The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to adequately demonstrate operability for the 4160 volt Standby Service Water Pump kerite cables through adequate testing and analysis in a continuously submerged environment. Furthermore, the environment for these continuously submerged cables exists because each of the two vaults that contain these cables (MH 20 and MH 21) has a design flaw, in that several other vaults gravity drain to them and the design of these vaults did not include a sump pump or other means for water to be removed or drained from them. The licensee has entered this into their corrective action program as CR-GGN-2009-01028. This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control of ensuring the availability, reliability, and capability of safety systems, and closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, because there was reasonable doubt on the continued operability of the Standby Service Water system. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution in that the licensee failed to implement Operating Experience directly communicated with a Generic Letter through changes to

station processes, procedures, and equipment. Inspection Report# : 2009006 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Maintenance Procedures to Maintain Drains on Safety Related Buildings The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V, involving the failure to properly clean and inspect the rooftop and associated water drainage systems of the safety-related diesel generator building. The inspectors identified loose, flexible roofing material that could have covered roof drains and result in loss of functionality for all of the standby diesel generators during a design basis heavy rainfall event. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-00429. This finding is more than minor since it affects the protection against external events attribute of mitigating system cornerstone. The roofing material and debris represented a degrading condition that if left uncorrected could have affected the availability, reliability, and capability of the standby diesel generators to respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding screened as potentially risk significant due to a flooding or severe weather initiating event, which then required a Phase 3 analysis. The Phase 3 analysis calculated a change in core damage frequency of 3.04E-8/yr, which represented very low safety significance. Inspection Report# : 2009002 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation for Standby Service Water Cooling Tower Drift Eliminators The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, involving a failure to perform an adequate operability evaluation. The inspectors identified non-conservatisms in the evaluation with regards to standby service water cooling tower drift rate, a failure to consider external events design basis impacts, and a failure to properly classify the condition as a substantially degraded, non-conforming condition, because it was subsequently determined that the deficiency could increase drift losses by a factor of ten. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-01222. This finding is more than minor because the failure to perform adequate operability evaluations, if left uncorrected, could become a more significant safety concern because the loss rates could become worse over time. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making because licensee personnel failed to use conservative assumptions and did not verify the validity of the underlying assumptions used in making safety-significant decisions. Inspection Report# : 2009002 (pdf) Barrier Integrity Significance: Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Fully Close a LPCS Manual Valve Resulted in Leakage of Water into the Condensate and Refueling Water Storage System The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1(a) involving a failure to implement the low pressure core spray system operating instruction correctly. On April 20, 2009, the site

was performing a low pressure core spray quarterly surveillance. During the test, the suppression pool level lowered approximately 0.8 inches, which equates to approximately 3600 gallons of water. Plant personnel investigated these anomalies and determined that the low pressure core spray pump had pressurized the condensate and refuelling water storage system due to a partially opened manual fill valve. This valve is a chain-fall operated valve and was approximately five turns open. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2009-02073. The finding was more than minor because it was associated with configuration control attribute of the reactor safety barrier integrity cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from the radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it only represents a degradation of the radiological barrier function provided for the auxiliary building. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources in that the operators did not have specific training in chain-fall type valve operation [H.2(b)] (Section 1R22). Inspection Report# : 2009003 (pdf) Significance: Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Enter a Limiting Condition for Operation for Primary Containment Isolation Valves The inspectors identified a Green noncited violation of Grand Gulf Nuclear Station Technical Specifications 3.6.1.3, for failure to enter a limiting condition for operation action statement for primary containment isolation valves. As a result, the limiting condition for operation action statement time was exceeded. The inspectors identified that surveillance test data for the residual heat removal Train A minimum flow valve was missing. The inspectors discovered that operations staff failed to properly review the work order for the valve work, and they had made an assumption the work order had been canceled. The licensee reviewed the identified issue for extent of condition and identified that in addition to a missed postmaintenance stroke test, they had also failed to enter the limiting condition for operation for two containment isolation valves. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-01069. This finding was more than minor since it affects the configuration control attribute of barrier integrity cornerstone, in that failing to properly test containment isolation valves could affect the assurance that physical design barriers that protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it did not represent an actual open pathway in the physical integrity of the containment system. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices, in that the operations shift supervisor and maintenance coordinator failed to perform proper self- and peer-checking and proper documentation of the completed work activity. Inspection Report# : 2009002 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : March 01, 2010

Grand Gulf 1 1Q/2010 Plant Inspection Findings Initiating Events Significance: Mar 27, 2010 Identified By: NRC Item Type: FIN Finding Failure to Follow Work Instructions Results in Loss of Buss and a Plant Transient (Section 4OA3) Green. The inspectors reviewed a self-revealing finding for a failure to follow work instructions resulting a in loss of 480V power to a bus and a plant transient. Specifically, contract workers were directed by work instructions to enter into a motor control center via its top cable tray to run cables to a spare breaker. Contrary to this, the contract electrical workers deviated from approved work instructions, causing a phase to ground short that tripped the motor control center and resulted in a plant transient. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01404. This finding is more than minor because it was associated with the initiating events cornerstone attribute of human performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. As a result, the issue was of very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the supervisor of the workers failed to ensure the contract workers followed the approved work instructions as required [H.4(c)]. (Section 4OA3.3) Inspection Report# : 2010002 (pdf) Significance: Mar 27, 2010 Identified By: NRC Item Type: FIN Finding Inadequate Actions in Response to a Steam Leak Result in an Automatic Reactor Scram (Section 4OA3) Green. The inspectors reviewed a self-revealing finding involving the failure of site management to ensure that adequate corrective actions were implemented to resolve the effects of a large steam leak in the turbine building. Specifically, the reactor experienced an automatic scram on low reactor water level due to the B reactor feed pump minimum flow valve failing open and a subsequent trip of the A reactor feed pump. The scram investigation determined that the minimum flow valve failed open due to condensation in a cable routing box. The condensation was caused by a large steam leak on the second stage moisture separator re-heater drain valve. Cable splices in the box were submerged in water and eventually caused those cables to short to ground. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01503. This finding is more than minor because it was associated with the initiating events cornerstone attribute of equipment performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. This is because the reactor feed pump B was able to restore reactor water level post scram. As a result, the issue was of very low safety significance (Green). The cause of this Enclosure finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to prioritize and thoroughly evaluate the extent of the cause of the water grounding sensitive electronic equipment in the vicinity of the steam leak [P.1(c)]. (Section 4OA3.4) Inspection Report# : 2010002 (pdf) Significance: Jun 23, 2009

Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Respond to Control Room Alarms in Accordance with Plant Procedures A self-revealing noncited violation of Technical Specification 5.4.1(a) was reviewed involving a failure to follow the fire alarm response procedure during a fire in the reactor feedwater pump area. The operators failed to investigate the source of a smoke alarm for an hour, allowing a fire to develop beyond the incipient stage before it was discovered. On November 17, 2008, a fire ignited in oil-soaked insulation on the reactor feedwater Pump B. After two weeks of plant operation following a refueling outage, during the November 17 shift turnover meeting, a fire alarm was received in the control room and was acknowledged by an operator. No notifications were made to the shift manager, and no operator or fire brigade member was dispatched. One hour after shift turnover, during normal operator rounds the turbine building operator discovered the fire in the reactor feedwater pump room. The fire brigade was dispatched to extinguish the fire. The licensee entered this condition in the corrective action program as condition report CR-GGN-2008-06584. The finding was more than minor because it was associated with the initiating events cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The finding was assessed by performing a bounding analysis using Appendix M of Inspection Manual Chapter 0609. The bounding analysis indicated that the change in core damage frequency would be 4.24 x 10-7 over a 1-year assessment period, indicating that this finding was of very low safety significance. This finding has a crosscutting aspect in the area of human performance with a work practices component for failure to use proper self checking techniques commensurate with the risk of the assigned task to ensure the work is performed safely because operators failed to use self checking techniques when acknowledging the reactor feedwater pump fire alarm [H.4(a)] (Section 4OA3). Inspection Report# : 2009003 (pdf) Mitigating Systems Significance: Mar 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Restore Control Room Air Conditioning Subsystem B to Operable Status Within the Required Time of 30 days (Section 1R07) Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.7.4 for failing to restore control room air conditioning subsystem B to operable status within the required time of 30 days. Specifically, between March 28, 2009 and June 25, 2009, the control room air conditioner subsystem B was inoperable due to the compressor capacity controller being set incorrectly. The deficiency initially revealed itself on May 14, 2009, when the air conditioner was unable to keep up with demand. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-3779. This finding is more than minor since it affects because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of safety related equipment. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, it does not represent an actual loss of a system safety function, it does not represent the actual loss of safety function of a single train for greater than its technical specification allowed outage time, it does not represent an actual loss of safety function of one or more non-technical specification of equipment designated as risk-significant per 10 CFR 50.65 for greater than 24 hours and it does not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making in that the operators did not utilize conservative assumptions to determine system operability [H.1(b)]. (Section 1R07.2). Inspection Report# : 2010002 (pdf) Significance: Dec 03, 2009

Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure Used to Vent the Reactor Core Isolation Cooling System Green. The team identified a noncited violation of Technical Specification Surveillance Requirement 3.5.3.1 because the licensee failed to establish an adequate procedure to demonstrate compliance with this surveillance requirement. This surveillance requires the licensee to Verify the RCIC System piping is filled with water from the pump discharge valve to the injection valve, every 31 days. To implement this requirement, the licensee vents the reactor core isolation cooling (RCIC) pump discharge leg from high point vents. However, the procedure failed to provide adequate qualitative or quantitative acceptance criteria to verify that the piping is maintained filled with water. This problem was previously documented in NRC Inspection Report 05000416/2007005 and entered into the corrective action program; however, the licensee failed to implement effective corrective actions. The failure of the licensee to effectively implement the surveillance requirement was a performance deficiency. This finding is more than minor because it affects the procedure quality 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 finding was determined to be of very low safety significance because it did not represent the loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. This finding has a crosscutting aspect in the corrective action program component of the problem identification and resolution area because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity [P.1(d)] (Section 4OA2.5a). Inspection Report# : 2009008 (pdf) Significance: Dec 03, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Oil-Impregnated Insulation on Pump Turbines Green. A noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was revealed on October 26, 2009, when potentially oil-impregnated insulation on the reactor core isolation cooling pump turbine began smoking during a monthly surveillance run. The reactor core isolation cooling pump turbine was shut down and the damaged insulation was removed and replaced. The maintenance technician indicated that the insulation was old and showed signs of possible oil impregnation; however, the licensee disposed of the insulation without performing an analysis. Oil-soaked insulation with a burn mark had previously been identified at a different location on the reactor core isolation cooling pump turbine on February 2, 2009. The licensee was unable to identify the source of the oil in either of these cases. Further, following a November 2008 fire in oil-soaked insulation on a reactor feed pump turbine, the licensee identified that the reactor core isolation cooling pump turbines were vulnerable to fire from similar causes due to a similar configuration. Corrective actions from the previous burnt insulation event and operating experience from the feed pump turbine both failed to prevent the October 26, 2009, smoke event. Burning of insulation on turbine-driven pump turbines and the potential for creating a fire is a significant condition adverse to quality. The failure of the licensee to determine the cause and to prevent recurrence of a significant condition adverse to quality was a performance deficiency. This finding is more than minor because it affects 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 determined that the performance deficiency to be of very low safety significance (Green) because it did not represent the loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding or severe weather initiating event. This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area because the licensee failed to implement and institutionalize operating experience on turbine insulation fires through changes to station processes, procedures, equipment, and training programs [P.2(b)] (Section 4OA2.5b). Inspection Report# : 2009008 (pdf) Significance: Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation

Failure to Monitor Performance of a Maintenance Rule Scoped System Green. The inspectors identified a Green noncited violation of 10 CFR Part 50.65(a)(2) involving the failure to adequately monitor the performance of a maintenance rule scoped system. The licensees maintenance rule program required evaluation of the area radiation monitoring system for classification as a maintenance rule (a)(1) system after three failures within eighteen months. The licensee had identified two functional failures of the residual heat removal heat exchanger A hatch radiation monitor in June and July 2008. The inspectors identified three other instances of functional failures on components that were used in plant emergency operating procedures and emergency preparedness procedures. These failures were not included in the licensees maintenance rule database. A total of five functional failures occurred in system components before the licensee considered evaluation of area radiation monitoring as a maintenance rule (a)(1) system in September 2009. The licensee entered this condition in the corrective action program as condition reports CR-GGN-2009-04853 and CR-GGN-2009-04857. The finding was more than minor because it was similar to Inspection Manual Chapter 0612, Appendix E, Example 7.d, in that equipment performance problems were such that effective control of performance or condition through appropriate preventive Maintenance Under (a)(2) could not be demonstrated. In addition, it affected the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was characterized under the significance determination process as having very low safety significance because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue does not reflect current licensee performance. (Section 1R12) Inspection Report# : 2009004 (pdf) Significance: Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Operator Response Times to Fires Green. The inspectors identified a Green non-cited violation of Technical Specification 5.4.1(a), for failure to ensure that operators can respond in timely manner to safe shutdown panels in the auxiliary building with a fire in the main control room. The inspectors reviewed a condition report associated with response times of operators to a fire in the protected area with Mississippi river at flood stage. The inspectors questioned the adequacy of response times for fire brigade members and the safe shutdown operator in the event of fire in the control room with the designated operators being outside the protected area. The licensee determined a time critical task would not have been completed due to the safe shutdown operator being outside the protected area. The licensee entered this condition in the corrective action program as condition report CR-GGN-2009-01416. The inspectors determined this finding to be more than minor since it affected the external events 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. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, it was determined that the finding screened as potentially risk significant due to external events and required the regional senior reactor analyst to perform a Phase 3 evaluation. The senior reactor analyst determined the likelihood that control room abandonment occurs while the safe shutdown operator is out of the protected area is 9.78E-8. The change in core damage frequency is lower than this value and small enough that large early release frequency is not required to be considered. Therefore the issue is (Green) of very low safety significance. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with corrective action program in that the licensee failed to perform an appropriate extent of condition when implementing corrective action associated with fire brigade response issue in 2008 [P.1(c)]. (Section 4OA2) Inspection Report# : 2009004 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Design Changes to Protect the Standby Service Water Slab

The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion III involving the failure to incorporate design changes required to limit dynamic loads on the standby service water basin slab. In 1997, the plant experienced damage to the standby service water basin slabs resulting from unanalyzed dynamic loads. During a standby service water system inspection on April 18, 2009, inspectors observed several different tire tracks on the seismically-designed concrete slab that covers and is integral to the safety-related standby service water basin. The inspectors also noted small placards attached to the basin slabs which prohibited moving vehicles on the slabs, and other signs requiring protective mats under any items placed on the slabs. Plant personnel evaluated the loading of the vehicle and determined that the load limits on the basin slab had not been exceeded. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2009 002087. The inspectors determined this finding affected the design control 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. In addition, the finding was more than minor because the failure to prevent dynamic loads on the standby service water basin slabs, if left uncorrected, could become more significant safety concern. Using the Manual Chapter of 0609, Significance Determination Process, Phase 1 Worksheet, this finding was determined to have very low safety significance, because it did not represent an actual loss of a safety function of the standby service water system. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue occurred several years ago and does not reflect current licensee performance (Section 1R04). Inspection Report# : 2009003 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation for Debris Left in the Condensate Storage Tank The inspectors identified a Green noncited violation of 10 CFR Part 50 Appendix B, Criterion V involving a failure to follow procedures which resulted in an inadequate operability evaluation. During the week of May 18, 2009, the site conducted debris removal in the condensate storage tank. This debris removal was necessary because of a failure to remove all debris in the condensate storage tank during their spring 2007 cleanup project. The licensee performed an operability evaluation for objects left in the condensate storage tank which stated that the high pressure core spray system and reactor core isolation cooling would remain operable for all postulated events. Upon review by the inspectors, the operability evaluation did not address several issues including objects left in the condensate storage tank and condensate system return flow to the condensate storage tank following a plant shutdown/scram. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2009-02815 and CR-GGN-2009-02837. This finding is more than minor because the failure to perform an adequate operability evaluation, if left uncorrected, could become a more significant safety concern. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to thoroughly identify all materials left in the condensate storage tank during their original operability determination [P.1(a)] (Section 1R15). Inspection Report# : 2009003 (pdf) Significance: Jun 23, 2009 Identified By: NRC Item Type: FIN Finding Failure to Perform a Timely Operability Evaluation Following the Discovery of a SSW Fan Failure Mechanism The inspectors identified a Green finding involving the failure to perform an operability determination after a new failure mechanism was discovered for standby service water Fan B. The inspectors were performing a follow up review of a condition report that detailed a trip of Division 1 standby service water Fan B. The fan had tripped on start up from the control room on December 31, 2007. The licensee had initially determined the trip was due to a faulty solid state trip device. Subsequent testing in failed to identify a problem with the trip device, and the apparent cause of the fan trip was attributed to reverse rotation of the fan. Operations personnel were not informed of this new

information as required by the corrective action program procedure. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2009-01726. This finding is more than minor because it was associated with the equipment performance attribute of the reactor safety mitigating systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the inspectors determined that the finding had very low safety significance (Green) since it did not represent an actual loss of a safety function of the standby service water cooling towers, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of human performance associated with work practices in that licensee personnel failed to apply procedural requirements to write a new condition report when new information was acquired related to the trip of the Division 1 standby service water Fan B [H.4(b)] (Section 4OA2). Inspection Report# : 2009003 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of Division II Standby Diesel Generator Jacket Water Heat Exchanger The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," for failure to comply with the licensees Generic Letter 89 13 program, which specifically states that cleaning of heat exchangers covered by this program is prohibited prior to performing an as-found thermal performance test. Specifically, in early 2006, the Division II Standby Diesel Generator (i.e. Emergency Diesel Generator) jacket water cooling heat exchanger was cleaned just prior to performing a five year thermal performance test. The licensee has entered this into their corrective action program as CR-GGN-2009-00904. This finding is more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of safety systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The inspectors reviewed the finding for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Non-conservative Bias in Instrumentation Used for Standby Service Water Leak Detection The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to establish adequate measures for the selection and review for suitability of equipment and processes that are essential to the safety-related functions of structures, systems and components. Specifically, the licensee failed to properly design for pulsation effects on flow rate instrumentation used for leak detection in the Standby Service Water system. This instrumentation is needed to meet licensee commitment 10 CFR Part 50, Appendix A, General Design Criterion 13, "Instrumentation and Control," to monitor trends in the ultimate heat sink basin inventory with the system in operation. The system was designed to detect a leakage rate of 1250 gallons per minute, and alarm in the control room at this leak rate, but due to design inadequacies in the instrumentation, the leak rate would have to exceed 3350 gallons per minute before activating the alarm. The licensee has entered this into their corrective action program as CR GGN 2009-00054. This finding was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1 3, "Screen for

More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The finding was reviewed for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Motor-Operated Valve Calculations Used Non-Conservative Inputs and Methodologies The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" for failing to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee used non conservative inputs or methodologies in calculating terminal voltages to safety related motor-operated valve motors that would be required to operate for mitigation of design bases events. The licensees electrical calculations used non conservative 50 percent locked rotor currents and neglected thermal overload resistance to determine the terminal voltages to safety related motor-operated valves which would predict higher terminal voltages than would actually exist. The calculated terminal voltages were direct design inputs into the applicable motor-operated valves mechanical thrust and torque calculations. The licensee has entered this issue into their corrective action program as CR GGN 2009 00985. This finding was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance of ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1 3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. This finding has a cross cutting aspect in the area of Problem Identification and Resolution, in that self assessments are of sufficient depth, are comprehensive, are appropriately objective, and are self critical. The licensee had conducted a Component Design Bases Assessment, LO GLO 2008 00044 in August 2008, and failed to adequately assess an identical finding identified at River Bend Station during their 2008 Component Design Bases Inspection. The licensee had determined that this issue was not applicable at Grand Gulf Nuclear Station. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions fo rReplacement of Safety-Related Batteries The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for failure to identify and correct a condition adverse to quality related to the seismic qualification of the Division III High Pressure Core Spray safety-related battery. Specifically, the licensee failed to identify an incorrectly installed end bracket after replacement of the Division III safety-related battery in 2002 using procedures, work instructions, and drawings that were supposed to have been corrected after this same issue was identified during a 1997 battery replacement activity. The licensee has entered this into their corrective action program as CR-GGN-2009-00830. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events for ensuring the availability, reliability, and capability of systems that respond to initiating events. Also, using Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, Section 1-3, "Screen for More than Minor - ROP," question 2, the finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was confirmed to not result in a loss of operability or functionality. The finding was

reviewed for cross-cutting aspects and none were identified. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Two Examples of a Failure to Meet 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," with two examples. Specifically, the team identified that the licensee failed to develop and implement adequate testing programs for Class 1E molded-case circuit breakers, and for the voltage and frequency response of the standby diesel generators that met design or vendor requirements and recommendations. In response, the licensee entered these examples in the corrective action program as CR GGN 2009 01024, and CR GGN-2009-01057. This finding was more than minor because it affected the mitigating systems cornerstone attribute of external events for ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, each example was determined to be of very low safety significance (Green) because they did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time, did not represent an actual loss of one or more risk-significant non-Technical Specification trains of equipment for greater than 24 hours, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding has a cross cutting aspect in the area of Problem Identification and Resolution, in that self assessments are of sufficient depth, are comprehensive, are appropriately objective, and are self critical. The licensee had conducted a Component Design Bases Assessment, LO-GLO-2008-00044 in August 2008, and failed to adequately assess an identical finding identified at River Bend Station during their 2008 Component Design Bases Inspection. The licensee had determined that this issue was not applicable at Grand Gulf Nuclear Station. Inspection Report# : 2009006 (pdf) Significance: Apr 02, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Control for Standby Service Water Pump Cables and Electrical Vaults The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to adequately demonstrate operability for the 4160 volt Standby Service Water Pump kerite cables through adequate testing and analysis in a continuously submerged environment. Furthermore, the environment for these continuously submerged cables exists because each of the two vaults that contain these cables (MH 20 and MH 21) has a design flaw, in that several other vaults gravity drain to them and the design of these vaults did not include a sump pump or other means for water to be removed or drained from them. The licensee has entered this into their corrective action program as CR-GGN-2009-01028. This finding is more than minor because it affected the mitigating systems cornerstone attribute of design control of ensuring the availability, reliability, and capability of safety systems, and closely parallels Inspection Manual Chapter 0612, Appendix E, Example 3.j, because there was reasonable doubt on the continued operability of the Standby Service Water system. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheets, the finding was determined to be of very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not affect external event mitigation. The inspectors determined that the finding has a crosscutting aspect in the area of Problem Identification and Resolution in that the licensee failed to implement Operating Experience directly communicated with a Generic Letter through changes to station processes, procedures, and equipment. Inspection Report# : 2009006 (pdf) Barrier Integrity

Significance: Mar 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Operator Failed to Move a Fuel Assembly in Accordance with Station Procedures Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1a when a fuel handling platform operator failed to move a fuel assembly in accordance with station procedures. Specifically, a new fuel assembly and the fuel handling platform mast were damaged when the platform was moved away from the fuel preparation machine prior to ensuring that the fuel assembly was clear of the machine. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01883. This finding is more than minor because the finding was associated with the human performance attribute of the barrier integrity cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that physical design barriers (i.e. fuel cladding) protect the public from radionuclide releases caused by accidents or events. The failure to follow the fuel handling procedures affected the cornerstones objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1-Initial Screening and Characterization of Findings, was used to evaluate the significance of the finding. Attachment 0609.04, Table 4a, was used to evaluate the impact of the finding on fuel clad integrity. Since the finding represented a fuel handling error that did not cause damage to fuel clad integrity, the finding was determined to be of very low safety significance (Green). The finding has a cross cutting aspect in the work practices component of the human performance area because the operator performing the fuel movement and the spotter providing oversight of the fuel movement failed to employ effective self and peer checking techniques such that fuel handling activities were performed safely [H.4.(a)]. (Section 4OA3.5) Inspection Report# : 2010002 (pdf) Significance: Jun 23, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Fully Close a LPCS Manual Valve Resulted in Leakage of Water into the Condensate and Refueling Water Storage System The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1(a) involving a failure to implement the low pressure core spray system operating instruction correctly. On April 20, 2009, the site was performing a low pressure core spray quarterly surveillance. During the test, the suppression pool level lowered approximately 0.8 inches, which equates to approximately 3600 gallons of water. Plant personnel investigated these anomalies and determined that the low pressure core spray pump had pressurized the condensate and refuelling water storage system due to a partially opened manual fill valve. This valve is a chain-fall operated valve and was approximately five turns open. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2009-02073. The finding was more than minor because it was associated with configuration control attribute of the reactor safety barrier integrity cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers would protect the public from the radionuclide releases caused by accident or events. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) since it only represents a degradation of the radiological barrier function provided for the auxiliary building. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources in that the operators did not have specific training in chain-fall type valve operation [H.2(b)] (Section 1R22). Inspection Report# : 2009003 (pdf) Emergency Preparedness

Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: N/A Dec 03, 2009 Identified By: NRC Item Type: FIN Finding Grand Gulf Nuclear Station 2009 Biennial PI&R Inspection Summary The inspectors concluded that the licensee was, in general, effective in identifying, evaluating, and resolving problems. Grand Gulf personnel were identifying and entering issues into the corrective action program at appropriately low thresholds as evidenced by a large number of condition reports issued; however, the team identified several minor deficiencies during walkdowns that had become masked to the employees due to the general lack of cleanliness in the plant. The team determined that the licensee generally screened issues appropriately for operability and reportability; however, five examples were identified where the licensee failed to perform an adequate operability determination. The team noted that issues were typically identified promptly and prioritized commensurate with their safety significance. Most root and apparent cause analyses appropriately considered extent of condition and previous occurrences. The team concluded that the corrective actions were generally identified and implemented promptly; however, the team noted several instances where corrective actions were not implemented or were cancelled. The team identified one cited violation of very low safety significance because of the licensees failure to restore compliance in a timely manner for a previously issued noncited violation. The team found that the licensee had established and was maintaining an environment at Grand Gulf where employees felt free to raise safety concerns without fear of retaliation. Inspection Report# : 2009008 (pdf) Last modified : May 26, 2010

Grand Gulf 1 2Q/2010 Plant Inspection Findings Initiating Events Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Work Instructions Results in Loss of Buss and a Plant Transient (Section 4OA3) Green. The inspectors reviewed a self-revealing finding for a failure to follow work instructions resulting a in loss of 480V power to a bus and a plant transient. Specifically, contract workers were directed by work instructions to enter into a motor control center via its top cable tray to run cables to a spare breaker. Contrary to this, the contract electrical workers deviated from approved work instructions, causing a phase to ground short that tripped the motor control center and resulted in a plant transient. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01404. This finding is more than minor because it was associated with the initiating events cornerstone attribute of human performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. As a result, the issue was of very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the supervisor of the workers failed to ensure the contract workers followed the approved work instructions as required [H.4(c)]. (Section 4OA3.3) Inspection Report# : 2010002 (pdf) Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Actions in Response to a Steam Leak Result in an Automatic Reactor Scram (Section 4OA3) Green. The inspectors reviewed a self-revealing finding involving the failure of site management to ensure that adequate corrective actions were implemented to resolve the effects of a large steam leak in the turbine building. Specifically, the reactor experienced an automatic scram on low reactor water level due to the B reactor feed pump minimum flow valve failing open and a subsequent trip of the A reactor feed pump. The scram investigation determined that the minimum flow valve failed open due to condensation in a cable routing box. The condensation was caused by a large steam leak on the second stage moisture separator re-heater drain valve. Cable splices in the box were submerged in water and eventually caused those cables to short to ground. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01503. This finding is more than minor because it was associated with the initiating events cornerstone attribute of equipment performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. This is because the reactor feed pump B was able to restore reactor water level post scram. As a result, the issue was of very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to prioritize and thoroughly evaluate the extent of the cause of the water grounding sensitive electronic equipment in the vicinity of the steam leak [P.1(c)]. (Section 4OA3.4) Inspection Report# : 2010002 (pdf)

Mitigating Systems Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Restore Control Room Air Conditioning Subsystem B to Operable Status Within the Required Time of 30 days (Section 1R07) Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.7.4 for failing to restore control room air conditioning subsystem B to operable status within the required time of 30 days. Specifically, between March 28, 2009 and June 25, 2009, the control room air conditioner subsystem B was inoperable due to the compressor capacity controller being set incorrectly. The deficiency initially revealed itself on May 14, 2009, when the air conditioner was unable to keep up with demand. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-3779. This finding is more than minor since it affects because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of safety related equipment. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, it does not represent an actual loss of a system safety function, it does not represent the actual loss of safety function of a single train for greater than its technical specification allowed outage time, it does not represent an actual loss of safety function of one or more non-technical specification of equipment designated as risk-significant per 10 CFR 50.65 for greater than 24 hours and it does not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making in that the operators did not utilize conservative assumptions to determine system operability [H.1(b)]. (Section 1R07.2). Inspection Report# : 2010002 (pdf) Significance: Dec 03, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure Used to Vent the Reactor Core Isolation Cooling System Green. The team identified a noncited violation of Technical Specification Surveillance Requirement 3.5.3.1 because the licensee failed to establish an adequate procedure to demonstrate compliance with this surveillance requirement. This surveillance requires the licensee to Verify the RCIC System piping is filled with water from the pump discharge valve to the injection valve, every 31 days. To implement this requirement, the licensee vents the reactor core isolation cooling (RCIC) pump discharge leg from high point vents. However, the procedure failed to provide adequate qualitative or quantitative acceptance criteria to verify that the piping is maintained filled with water. This problem was previously documented in NRC Inspection Report 05000416/2007005 and entered into the corrective action program; however, the licensee failed to implement effective corrective actions. The failure of the licensee to effectively implement the surveillance requirement was a performance deficiency. This finding is more than minor because it affects the procedure quality 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 finding was determined to be of very low safety significance because it did not represent the loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. This finding has a crosscutting aspect in the corrective action program component of the problem identification and resolution area because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity [P.1(d)] (Section 4OA2.5a). Inspection Report# : 2009008 (pdf) Significance: Dec 03, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Oil-Impregnated Insulation on Pump Turbines

Green. A noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was revealed on October 26, 2009, when potentially oil-impregnated insulation on the reactor core isolation cooling pump turbine began smoking during a monthly surveillance run. The reactor core isolation cooling pump turbine was shut down and the damaged insulation was removed and replaced. The maintenance technician indicated that the insulation was old and showed signs of possible oil impregnation; however, the licensee disposed of the insulation without performing an analysis. Oil-soaked insulation with a burn mark had previously been identified at a different location on the reactor core isolation cooling pump turbine on February 2, 2009. The licensee was unable to identify the source of the oil in either of these cases. Further, following a November 2008 fire in oil-soaked insulation on a reactor feed pump turbine, the licensee identified that the reactor core isolation cooling pump turbines were vulnerable to fire from similar causes due to a similar configuration. Corrective actions from the previous burnt insulation event and operating experience from the feed pump turbine both failed to prevent the October 26, 2009, smoke event. Burning of insulation on turbine-driven pump turbines and the potential for creating a fire is a significant condition adverse to quality. The failure of the licensee to determine the cause and to prevent recurrence of a significant condition adverse to quality was a performance deficiency. This finding is more than minor because it affects 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 determined that the performance deficiency to be of very low safety significance (Green) because it did not represent the loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding or severe weather initiating event. This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area because the licensee failed to implement and institutionalize operating experience on turbine insulation fires through changes to station processes, procedures, equipment, and training programs [P.2(b)] (Section 4OA2.5b). Inspection Report# : 2009008 (pdf) Significance: Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor Performance of a Maintenance Rule Scoped System Green. The inspectors identified a Green noncited violation of 10 CFR Part 50.65(a)(2) involving the failure to adequately monitor the performance of a maintenance rule scoped system. The licensees maintenance rule program required evaluation of the area radiation monitoring system for classification as a maintenance rule (a)(1) system after three failures within eighteen months. The licensee had identified two functional failures of the residual heat removal heat exchanger A hatch radiation monitor in June and July 2008. The inspectors identified three other instances of functional failures on components that were used in plant emergency operating procedures and emergency preparedness procedures. These failures were not included in the licensees maintenance rule database. A total of five functional failures occurred in system components before the licensee considered evaluation of area radiation monitoring as a maintenance rule (a)(1) system in September 2009. The licensee entered this condition in the corrective action program as condition reports CR-GGN-2009-04853 and CR-GGN-2009-04857. The finding was more than minor because it was similar to Inspection Manual Chapter 0612, Appendix E, Example 7.d, in that equipment performance problems were such that effective control of performance or condition through appropriate preventive Maintenance Under (a)(2) could not be demonstrated. In addition, it affected the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was characterized under the significance determination process as having very low safety significance because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. There is no crosscutting aspect associated with this performance deficiency since the cause of this issue does not reflect current licensee performance. (Section 1R12) Inspection Report# : 2009004 (pdf) Significance: Sep 23, 2009 Identified By: NRC Item Type: NCV NonCited Violation

Failure to Maintain Operator Response Times to Fires Green. The inspectors identified a Green non-cited violation of Technical Specification 5.4.1(a), for failure to ensure that operators can respond in timely manner to safe shutdown panels in the auxiliary building with a fire in the main control room. The inspectors reviewed a condition report associated with response times of operators to a fire in the protected area with Mississippi river at flood stage. The inspectors questioned the adequacy of response times for fire brigade members and the safe shutdown operator in the event of fire in the control room with the designated operators being outside the protected area. The licensee determined a time critical task would not have been completed due to the safe shutdown operator being outside the protected area. The licensee entered this condition in the corrective action program as condition report CR-GGN-2009-01416. The inspectors determined this finding to be more than minor since it affected the external events 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. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 worksheet, it was determined that the finding screened as potentially risk significant due to external events and required the regional senior reactor analyst to perform a Phase 3 evaluation. The senior reactor analyst determined the likelihood that control room abandonment occurs while the safe shutdown operator is out of the protected area is 9.78E-8. The change in core damage frequency is lower than this value and small enough that large early release frequency is not required to be considered. Therefore the issue is (Green) of very low safety significance. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with corrective action program in that the licensee failed to perform an appropriate extent of condition when implementing corrective action associated with fire brigade response issue in 2008 [P.1(c)]. (Section 4OA2) Inspection Report# : 2009004 (pdf) Barrier Integrity Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Operator Failed to Move a Fuel Assembly in Accordance with Station Procedures Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1a when a fuel handling platform operator failed to move a fuel assembly in accordance with station procedures. Specifically, a new fuel assembly and the fuel handling platform mast were damaged when the platform was moved away from the fuel preparation machine prior to ensuring that the fuel assembly was clear of the machine. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01883. This finding is more than minor because the finding was associated with the human performance attribute of the barrier integrity cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that physical design barriers (i.e. fuel cladding) protect the public from radionuclide releases caused by accidents or events. The failure to follow the fuel handling procedures affected the cornerstones objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1-Initial Screening and Characterization of Findings, was used to evaluate the significance of the finding. Attachment 0609.04, Table 4a, was used to evaluate the impact of the finding on fuel clad integrity. Since the finding represented a fuel handling error that did not cause damage to fuel clad integrity, the finding was determined to be of very low safety significance (Green). The finding has a cross cutting aspect in the work practices component of the human performance area because the operator performing the fuel movement and the spotter providing oversight of the fuel movement failed to employ effective self and peer checking techniques such that fuel handling activities were performed safely [H.4.(a)]. (Section 4OA3.5) Inspection Report# : 2010002 (pdf) Emergency Preparedness

Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: N/A Dec 03, 2009 Identified By: NRC Item Type: FIN Finding Grand Gulf Nuclear Station 2009 Biennial PI&R Inspection Summary The inspectors concluded that the licensee was, in general, effective in identifying, evaluating, and resolving problems. Grand Gulf personnel were identifying and entering issues into the corrective action program at appropriately low thresholds as evidenced by a large number of condition reports issued; however, the team identified several minor deficiencies during walkdowns that had become masked to the employees due to the general lack of cleanliness in the plant. The team determined that the licensee generally screened issues appropriately for operability and reportability; however, five examples were identified where the licensee failed to perform an adequate operability determination. The team noted that issues were typically identified promptly and prioritized commensurate with their safety significance. Most root and apparent cause analyses appropriately considered extent of condition and previous occurrences. The team concluded that the corrective actions were generally identified and implemented promptly; however, the team noted several instances where corrective actions were not implemented or were cancelled. The team found that the licensee had established and was maintaining an environment at Grand Gulf where employees felt free to raise safety concerns without fear of retaliation. Inspection Report# : 2009008 (pdf) Last modified : September 02, 2010

Grand Gulf 1 3Q/2010 Plant Inspection Findings Initiating Events Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Work Instructions Results in Loss of Buss and a Plant Transient (Section 4OA3) Green. The inspectors reviewed a self-revealing finding for a failure to follow work instructions resulting a in loss of 480V power to a bus and a plant transient. Specifically, contract workers were directed by work instructions to enter into a motor control center via its top cable tray to run cables to a spare breaker. Contrary to this, the contract electrical workers deviated from approved work instructions, causing a phase to ground short that tripped the motor control center and resulted in a plant transient. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01404. This finding is more than minor because it was associated with the initiating events cornerstone attribute of human performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. As a result, the issue was of very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the supervisor of the workers failed to ensure the contract workers followed the approved work instructions as required [H.4(c)]. (Section 4OA3.3) Inspection Report# : 2010002 (pdf) Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Actions in Response to a Steam Leak Result in an Automatic Reactor Scram (Section 4OA3) Green. The inspectors reviewed a self-revealing finding involving the failure of site management to ensure that adequate corrective actions were implemented to resolve the effects of a large steam leak in the turbine building. Specifically, the reactor experienced an automatic scram on low reactor water level due to the B reactor feed pump minimum flow valve failing open and a subsequent trip of the A reactor feed pump. The scram investigation determined that the minimum flow valve failed open due to condensation in a cable routing box. The condensation was caused by a large steam leak on the second stage moisture separator re-heater drain valve. Cable splices in the box were submerged in water and eventually caused those cables to short to ground. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01503. This finding is more than minor because it was associated with the initiating events cornerstone attribute of equipment performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. This is because the reactor feed pump B was able to restore reactor water level post scram. As a result, the issue was of very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to prioritize and thoroughly evaluate the extent of the cause of the water grounding sensitive electronic equipment in the vicinity of the steam leak [P.1(c)]. (Section 4OA3.4) Inspection Report# : 2010002 (pdf)

Mitigating Systems Significance: Jun 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment for Emergent Work Activities

  • Green. The inspectors identified a noncited violation of 10 CFR 50.65 a(4) for failure to perform adequate risk assessments prior to flushing the reactor heat removal systems suction piping and filling and venting of the alternate decay heat removal system. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2010 02553.

This finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone, adversely affecting the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors evaluated the finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. Using Appendix K, the inspectors determined that the finding has a very low safety significance because the finding was related only to the performance of risk management actions and did not exceed the threshold for core damage probability and large early release probability. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources, because the licensee failed to provide adequate training on the implementation of the new risk management procedure [H.2(b)] (Section 1R13.b). Inspection Report# : 2010003 (pdf) Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Restore Control Room Air Conditioning Subsystem B to Operable Status Within the Required Time of 30 days (Section 1R07) Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.7.4 for failing to restore control room air conditioning subsystem B to operable status within the required time of 30 days. Specifically, between March 28, 2009 and June 25, 2009, the control room air conditioner subsystem B was inoperable due to the compressor capacity controller being set incorrectly. The deficiency initially revealed itself on May 14, 2009, when the air conditioner was unable to keep up with demand. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-3779. This finding is more than minor since it affects because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of safety related equipment. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, it does not represent an actual loss of a system safety function, it does not represent the actual loss of safety function of a single train for greater than its technical specification allowed outage time, it does not represent an actual loss of safety function of one or more non-technical specification of equipment designated as risk-significant per 10 CFR 50.65 for greater than 24 hours and it does not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making in that the operators did not utilize conservative assumptions to determine system operability [H.1(b)]. (Section 1R07.2). Inspection Report# : 2010002 (pdf) Significance: Dec 03, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure Used to Vent the Reactor Core Isolation Cooling System Green. The team identified a noncited violation of Technical Specification Surveillance Requirement 3.5.3.1 because

the licensee failed to establish an adequate procedure to demonstrate compliance with this surveillance requirement. This surveillance requires the licensee to Verify the RCIC System piping is filled with water from the pump discharge valve to the injection valve, every 31 days. To implement this requirement, the licensee vents the reactor core isolation cooling (RCIC) pump discharge leg from high point vents. However, the procedure failed to provide adequate qualitative or quantitative acceptance criteria to verify that the piping is maintained filled with water. This problem was previously documented in NRC Inspection Report 05000416/2007005 and entered into the corrective action program; however, the licensee failed to implement effective corrective actions. The failure of the licensee to effectively implement the surveillance requirement was a performance deficiency. This finding is more than minor because it affects the procedure quality 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 finding was determined to be of very low safety significance because it did not represent the loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. This finding has a crosscutting aspect in the corrective action program component of the problem identification and resolution area because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity [P.1(d)] (Section 4OA2.5a). Inspection Report# : 2009008 (pdf) Significance: Dec 03, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Oil-Impregnated Insulation on Pump Turbines Green. A noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was revealed on October 26, 2009, when potentially oil-impregnated insulation on the reactor core isolation cooling pump turbine began smoking during a monthly surveillance run. The reactor core isolation cooling pump turbine was shut down and the damaged insulation was removed and replaced. The maintenance technician indicated that the insulation was old and showed signs of possible oil impregnation; however, the licensee disposed of the insulation without performing an analysis. Oil-soaked insulation with a burn mark had previously been identified at a different location on the reactor core isolation cooling pump turbine on February 2, 2009. The licensee was unable to identify the source of the oil in either of these cases. Further, following a November 2008 fire in oil-soaked insulation on a reactor feed pump turbine, the licensee identified that the reactor core isolation cooling pump turbines were vulnerable to fire from similar causes due to a similar configuration. Corrective actions from the previous burnt insulation event and operating experience from the feed pump turbine both failed to prevent the October 26, 2009, smoke event. Burning of insulation on turbine-driven pump turbines and the potential for creating a fire is a significant condition adverse to quality. The failure of the licensee to determine the cause and to prevent recurrence of a significant condition adverse to quality was a performance deficiency. This finding is more than minor because it affects 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 determined that the performance deficiency to be of very low safety significance (Green) because it did not represent the loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding or severe weather initiating event. This finding has a cross-cutting aspect in the operating experience component of the problem identification and resolution area because the licensee failed to implement and institutionalize operating experience on turbine insulation fires through changes to station processes, procedures, equipment, and training programs [P.2(b)] (Section 4OA2.5b). Inspection Report# : 2009008 (pdf) Barrier Integrity Significance: Jun 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation

Failure to Follow Procedure Results in Loss of Decay Heat Removal to the Spent Fuel Pool Green. The inspectors reviewed a self-revealing noncited violation of Technical Specifications 5.4.1(a), involving a loss of decay heat removal in the spent fuel pool due to station operators failing to follow the fuel pool cooling and cleanup system operating instruction. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2010 02172. This finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding has a very low safety significance because it only represented a loss of spent fuel pool cooling that would not preclude restoration of cooling to the spent fuel pool prior to pool boiling. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices, because licensee personnel failed to use adequate self- and peer-checking techniques to remove the filter/demineralizer from service [H.4(a)] (Section 1R15.b). Inspection Report# : 2010003 (pdf) Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Operator Failed to Move a Fuel Assembly in Accordance with Station Procedures Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1a when a fuel handling platform operator failed to move a fuel assembly in accordance with station procedures. Specifically, a new fuel assembly and the fuel handling platform mast were damaged when the platform was moved away from the fuel preparation machine prior to ensuring that the fuel assembly was clear of the machine. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01883. This finding is more than minor because the finding was associated with the human performance attribute of the barrier integrity cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that physical design barriers (i.e. fuel cladding) protect the public from radionuclide releases caused by accidents or events. The failure to follow the fuel handling procedures affected the cornerstones objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1-Initial Screening and Characterization of Findings, was used to evaluate the significance of the finding. Attachment 0609.04, Table 4a, was used to evaluate the impact of the finding on fuel clad integrity. Since the finding represented a fuel handling error that did not cause damage to fuel clad integrity, the finding was determined to be of very low safety significance (Green). The finding has a cross cutting aspect in the work practices component of the human performance area because the operator performing the fuel movement and the spotter providing oversight of the fuel movement failed to employ effective self and peer checking techniques such that fuel handling activities were performed safely [H.4.(a)]. (Section 4OA3.5) Inspection Report# : 2010002 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: N/A Dec 03, 2009 Identified By: NRC Item Type: FIN Finding Grand Gulf Nuclear Station 2009 Biennial PI&R Inspection Summary The inspectors concluded that the licensee was, in general, effective in identifying, evaluating, and resolving problems. Grand Gulf personnel were identifying and entering issues into the corrective action program at appropriately low thresholds as evidenced by a large number of condition reports issued; however, the team identified several minor deficiencies during walkdowns that had become masked to the employees due to the general lack of cleanliness in the plant. The team determined that the licensee generally screened issues appropriately for operability and reportability; however, five examples were identified where the licensee failed to perform an adequate operability determination. The team noted that issues were typically identified promptly and prioritized commensurate with their safety significance. Most root and apparent cause analyses appropriately considered extent of condition and previous occurrences. The team concluded that the corrective actions were generally identified and implemented promptly; however, the team noted several instances where corrective actions were not implemented or were cancelled. The team found that the licensee had established and was maintaining an environment at Grand Gulf where employees felt free to raise safety concerns without fear of retaliation. Inspection Report# : 2009008 (pdf) Last modified : November 29, 2010

Grand Gulf 1 4Q/2010 Plant Inspection Findings Initiating Events Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Switchyard Battery Replacement Green. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for failure to properly assess the risk impact of maintenance on the switchyard batteries. Specifically, plant personnel evaluated the work as light switchyard work when it should have been evaluated as heavy equipment, which increases the likelihood of a loss of offsite power transient. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06668. The finding was more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Event (IE) Cornerstone. Because the finding affects the licensees assessment of risk associated with performing maintenance activities, IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, directs significance determination via the use of IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Flowchart 1 of Appendix K, the significance of this finding was determined to be of very low safety significance (Green), because the calculated Incremental Core Damage Probability Deficit (2E-8) was not greater than 1.0E-6. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to use a systematic decision making process using available risk assessment guidance and did not obtain interdisciplinary input on an important risk management decision [H.1(a)] (Section 1R13). Inspection Report# : 2010004 (pdf) Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: FIN Finding Failure to Follow Work Instructions Results in Loss of Buss and a Plant Transient (Section 4OA3) Green. The inspectors reviewed a self-revealing finding for a failure to follow work instructions resulting a in loss of 480V power to a bus and a plant transient. Specifically, contract workers were directed by work instructions to enter into a motor control center via its top cable tray to run cables to a spare breaker. Contrary to this, the contract electrical workers deviated from approved work instructions, causing a phase to ground short that tripped the motor control center and resulted in a plant transient. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01404. This finding is more than minor because it was associated with the initiating events cornerstone attribute of human performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. As a result, the issue was of very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices because the supervisor of the workers failed to ensure the contract workers followed the approved work instructions as required [H.4(c)]. (Section 4OA3.3) Inspection Report# : 2010002 (pdf) Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: FIN Finding Inadequate Actions in Response to a Steam Leak Result in an Automatic Reactor Scram (Section 4OA3) Green. The inspectors reviewed a self-revealing finding involving the failure of site management to ensure that

adequate corrective actions were implemented to resolve the effects of a large steam leak in the turbine building. Specifically, the reactor experienced an automatic scram on low reactor water level due to the B reactor feed pump minimum flow valve failing open and a subsequent trip of the A reactor feed pump. The scram investigation determined that the minimum flow valve failed open due to condensation in a cable routing box. The condensation was caused by a large steam leak on the second stage moisture separator re-heater drain valve. Cable splices in the box were submerged in water and eventually caused those cables to short to ground. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01503. This finding is more than minor because it was associated with the initiating events cornerstone attribute of equipment performance, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during shutdown, as well as during power operations. Using the Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. This is because the reactor feed pump B was able to restore reactor water level post scram. As a result, the issue was of very low safety significance (Green). The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to prioritize and thoroughly evaluate the extent of the cause of the water grounding sensitive electronic equipment in the vicinity of the steam leak [P.1(c)]. (Section 4OA3.4) Inspection Report# : 2010002 (pdf) Mitigating Systems Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Degraded Fire Door Barrier Protecting the Safeguards Switchgear Rooms Green. The inspectors identified a Green noncited violation of Facility Operating License Condition 2.C(41) involving the failure to ensure that fire barriers protecting safety-related areas were functional during monthly fire barrier inspections. The inspectors identified that fire door OC211, crossover door between division 1 and 2 switchgear rooms, was missing 5 screws in the divider overlap and there was a three inch crack in the door on the bottom left side. The Fire Hazards Analysis Report, Section 9A.2.4 defines fire doors as a fire barrier, and Section 9A.5.7 and 9A.5.8, Fire Area 7 and Fire Area 8, respectively, describe the electrical switchgear rooms as having 3-hour fire rated barriers. Operations initiated an hourly fire watch for the non-functional door per the technical requirements manual. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2010-05541. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety Mitigating Systems (MS) Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire confinement category. The inspectors assigned a low degradation rating because the cracks did not create an actual hole through the door. The inspectors concluded that the finding was of very low safety significance (Green) because the degraded barrier was expected to maintain nearly the same level of effectiveness and reliability had the degradation not been present, and there were no fire ignition sources or combustible materials in the area that would subject the barrier to direct flame impingement. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources, because plant personnel failed to adequately evaluate and provide proper maintenance for degrading fire doors [H.2(d)] (Section 1R05) Inspection Report# : 2010004 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Remove Foreign Material from the Control Room Air Conditioning Systems Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, for the failure to remove foreign material from the control room air conditioning oil and Freon subsystems. The pencil strainer on

the compressor was found to be 90 percent clogged by foreign material. Plant personnel cleaned the pencil strainer, but placed the CRAC B system back in service without cleaning the oil and Freon subsystems which resulted in the CRAC B system becoming inoperable two weeks later. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-04839. The finding was more than minor because it was associated with the equipment performance attribute of the reactor safety Mitigating Systems (MS) Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance since it did not result in a loss of system safety function. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to appropriately address the foreign material in the control room air conditioning subsystems [P.1(d)] (Section 1R15). Inspection Report# : 2010004 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Implement an Adequate Structural Monitoring Program Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B Criterion V for failure to perform required inspection of safety-related plant structures. Specifically, the inspectors found inspections that had been only partially performed and some areas that had not been documented as inspected. Subsequent walkdowns identified several deficiencies including concrete cracks and spalling, deficient coatings, rusted tanks and exposed rebar. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06871. The finding is greater than minor because it is associated with the Mitigating Systems (MS) Cornerstone attribute of protection against external events and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance since it did not represent a loss of system safety function, an actual loss of safety function of a single train for greater than its TS allowed outage time, or screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of human performance, associated with the resources component, in that the licensee failed to accurately document and manage the structural inspections [H.2(c)] (Section 40A2). Inspection Report# : 2010004 (pdf) Significance: Jun 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment for Emergent Work Activities

  • Green. The inspectors identified a noncited violation of 10 CFR 50.65 a(4) for failure to perform adequate risk assessments prior to flushing the reactor heat removal systems suction piping and filling and venting of the alternate decay heat removal system. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2010 02553.

This finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone, adversely affecting the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors evaluated the finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. Using Appendix K, the inspectors determined that the finding has a very low safety significance because the finding was related only to the performance of risk management actions and did not exceed the threshold for core damage probability and large early release probability. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources, because the licensee failed to provide adequate training on the implementation of the new risk management procedure [H.2(b)] (Section

1R13.b). Inspection Report# : 2010003 (pdf) Significance: Mar 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Restore Control Room Air Conditioning Subsystem B to Operable Status Within the Required Time of 30 days (Section 1R07) Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.7.4 for failing to restore control room air conditioning subsystem B to operable status within the required time of 30 days. Specifically, between March 28, 2009 and June 25, 2009, the control room air conditioner subsystem B was inoperable due to the compressor capacity controller being set incorrectly. The deficiency initially revealed itself on May 14, 2009, when the air conditioner was unable to keep up with demand. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2009-3779. This finding is more than minor since it affects because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of safety related equipment. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, it does not represent an actual loss of a system safety function, it does not represent the actual loss of safety function of a single train for greater than its technical specification allowed outage time, it does not represent an actual loss of safety function of one or more non-technical specification of equipment designated as risk-significant per 10 CFR 50.65 for greater than 24 hours and it does not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision making in that the operators did not utilize conservative assumptions to determine system operability [H.1(b)]. (Section 1R07.2). Inspection Report# : 2010002 (pdf) Barrier Integrity Significance: Jun 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Loss of Decay Heat Removal to the Spent Fuel Pool Green. The inspectors reviewed a self-revealing noncited violation of Technical Specifications 5.4.1(a), involving a loss of decay heat removal in the spent fuel pool due to station operators failing to follow the fuel pool cooling and cleanup system operating instruction. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2010 02172. This finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding has a very low safety significance because it only represented a loss of spent fuel pool cooling that would not preclude restoration of cooling to the spent fuel pool prior to pool boiling. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices, because licensee personnel failed to use adequate self- and peer-checking techniques to remove the filter/demineralizer from service [H.4(a)] (Section 1R15.b). Inspection Report# : 2010003 (pdf) Significance: Mar 27, 2010

Identified By: Self-Revealing Item Type: NCV NonCited Violation Operator Failed to Move a Fuel Assembly in Accordance with Station Procedures Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1a when a fuel handling platform operator failed to move a fuel assembly in accordance with station procedures. Specifically, a new fuel assembly and the fuel handling platform mast were damaged when the platform was moved away from the fuel preparation machine prior to ensuring that the fuel assembly was clear of the machine. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-01883. This finding is more than minor because the finding was associated with the human performance attribute of the barrier integrity cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that physical design barriers (i.e. fuel cladding) protect the public from radionuclide releases caused by accidents or events. The failure to follow the fuel handling procedures affected the cornerstones objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1-Initial Screening and Characterization of Findings, was used to evaluate the significance of the finding. Attachment 0609.04, Table 4a, was used to evaluate the impact of the finding on fuel clad integrity. Since the finding represented a fuel handling error that did not cause damage to fuel clad integrity, the finding was determined to be of very low safety significance (Green). The finding has a cross cutting aspect in the work practices component of the human performance area because the operator performing the fuel movement and the spotter providing oversight of the fuel movement failed to employ effective self and peer checking techniques such that fuel handling activities were performed safely [H.4.(a)]. (Section 4OA3.5) Inspection Report# : 2010002 (pdf) Emergency Preparedness Significance: Sep 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Primary Meteorological Tower Inoperable Without Compensatory Actions in Place Green. A self-revealing non-cited violation of 10 CFR 50.47(b)(8), was identified when the Grand Gulf Nuclear Station Primary Meteorological Tower was rendered inoperable without compensatory actions from July 6 through July 27, 2010. The primary meteorological tower was declared inoperable by operations for maintenance to perform surveillance and preventative maintenance activities. The technicians did not finish the surveillance due to problems with data points exceeding allowable tolerance limits, and left the tower with the 10 and 50 meter instruments lowered to the ground. Inaccurate meteorological data continued to be displayed in the plant computer system. During the subsequent night shift, the control room supervisor inadvertently closed out the limiting condition of operations for the primary meteorological tower being out of service prior to the tower being returned to an operable condition. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2010-05748. The finding was more than minor because it was associated with the Facilities and Equipment attribute of the Emergency Preparedness (EP) Cornerstone and adversely affected the cornerstone objective of ensuring the capability to implement adequate measures to protect public health and safety in the event of a radiological emergency. Specifically, from July 6 through July 27, 2010, key emergency response members could not have accurately performed their assigned emergency notification and dose assessment functions, with an absence of compensatory measures. In accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1 and the examples contained in section 4.8 of the same document, the inspectors determined the finding to be of very low safety significance (Green) because the performance deficiency was a failure to comply with NRC regulations, the deficiency was associated with a non-risk significant planning standard as defined in MC0609 Appendix B, and it did not represent a functional failure of the planning standard. The cause of this finding has a crosscutting aspect in the area of human performance associated with work control, because the maintenance and operations department failed to appropriately communicate and coordinate work activities on the primary meteorological tower. [H.3(b)] (Section 1R19). Inspection Report# : 2010004 (pdf)

Occupational Radiation Safety Significance: Sep 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow the Radiological Protection Job Coverage Procedure Green. The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1.a for a procedure violation. Radiation Work Permit 20101704 covered work on Valve 1G33F253 in the reactor water cleanup room. Work on this valve was conducted over a 6-day period, May 6 through 11, 2010, and in that time, three personnel contaminations occurred. Appropriate protective clothing was not assigned by the job coverage technician and contributed to the three personnel contaminations and radioactive intake by one of the workers of 62 mrem. The failure to assign appropriate protective clothing during radiological work is a performance deficiency. The finding is greater than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process (exposure control), and affected the cornerstone objective, in that it resulted in an individual receiving unplanned dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work practices, because the radiation protection technician covering the job did not use risk insights or take the job site condition into consideration when assigning protective clothing for radiological work [H.3(a)] (2RSO4). Inspection Report# : 2010004 (pdf) Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : March 03, 2011

Grand Gulf 1 1Q/2011 Plant Inspection Findings Initiating Events Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Shutdown Procedure Resulting in Power and Level Oscillations in the Reactor Green. The inspectors identified a green, noncited violation of 10 CFR 50 Appendix B, Criterion V, for an inadequate shutdown procedure resulting in power and level oscillations in the reactor. The revised procedure failed to require the plant to be placed in startup feedwater level control during low power operations. In addition, the operators performed shutdown training on the old procedure. The performance deficiency was self-revealing, however the inspectors added significant value by identifying inadequate condition report classification, causal evaluation, and corrective actions. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-05140. The finding is more than minor because it was associated with the initiating events cornerstone attribute of procedure quality and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the inspectors determined that the finding impacted both the initiating events and mitigating systems cornerstones. The inspectors determined that the initiating event cornerstone best reflected the dominant risk of the finding. The finding was determined to be of very low safety significance (Green) because the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision-making, because station management failed to use conservative assumptions to demonstrate that the change to the shutdown operating procedure was safe prior to proceeding [H.1(b)]. (Section 4OA2) Inspection Report# : 2010005 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Switchyard Battery Replacement Green. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for failure to properly assess the risk impact of maintenance on the switchyard batteries. Specifically, plant personnel evaluated the work as light switchyard work when it should have been evaluated as heavy equipment, which increases the likelihood of a loss of offsite power transient. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06668. The finding was more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Event (IE) Cornerstone. Because the finding affects the licensees assessment of risk associated with performing maintenance activities, IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, directs significance determination via the use of IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Flowchart 1 of Appendix K, the significance of this finding was determined to be of very low safety significance (Green), because the calculated Incremental Core Damage Probability Deficit (2E-8) was not greater than 1.0E-6. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to use a systematic decision making process using available risk assessment guidance and did not obtain interdisciplinary input on an important risk management decision [H.1(a)] (Section 1R13). Inspection Report# : 2010004 (pdf)

Mitigating Systems Significance: SL-IV Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update Available Low Pressure Coolant Injection Loops in the Updated Final Safety Analysis Report SLIV. Inspectors identified a noncited violation of 10 CFR 50.71(e)(4), which requires the final safety analysis report be updated, at intervals not exceeding 24 months, to reflect changes made in the facility or procedures described in the final safety analysis report. Licensee personnel failed to update the original revision of the final safety analysis report to reflect the actual number of low pressure coolant injection loops available for automatic initiation during shutdown cooling operations in Mode 3. The licensee plans to update the final safety analysis report at the next scheduled revision. This finding was entered into the licensees corrective action program as condition report CR-GGN-2011-01631. The failure of licensing personnel to update the final safety analysis report to reflect the available low pressure coolant injection loops for automatic initiation during shutdown cooling operations in Mode 3 was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy, dated September 30, 2010, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate Maintenance Effectiveness of Train B Control Room Air Conditioner Green. The inspectors identified a noncited violation of 10 CFR Part 50.65(a)(2) for the licensees failure to demonstrate that the performance of the train B control room air conditioner was being effectively controlled through the performance of appropriate preventive maintenance. Engineering did not properly evaluate maintenance rule functional failures resulting in the system remaining in an a(2) status instead of an a(1) status. As corrective action, the train B control room air conditioner was moved into an a(1) status. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2011-01623. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. As corrective action, the train B control room air conditioner was moved into an (a)(1) status. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because licensee personnel failed to make appropriate safety-significant or risk-significant decisions to address the multiple failures of the train B control room air conditioner compressor. [H.1(a)] (Section 1R12.b.2) Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Control Room Air Conditioner Compressor Tripping Due to Low Oil Pressure Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after the licensee failed to determine the cause and prevent recurrence of a significant condition adverse to quality associated with the train B control room air conditioner compressor tripping due to low oil pressure. Specifically, on December 13, 2010, the train B control room air conditioner compressor tripped on low oil pressure

after the licensee had performed a root cause analysis to identify the cause and prevent recurrence of a similar compressor trip on October 14, 2010. As immediate corrective action, the licensee installed an inline suction filter. No additional failures have occurred since its installation. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2010-07315. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, the inspectors determined that a Phase 2 analysis was required because the finding represented a loss of system safety function. The plant-specific risk informed notebook does not include the evaluation of risk caused by the loss of cooling to the main control room. Therefore, the senior reactor analyst conducted a Phase 3 analysis. Based on the bounding analysis, the analyst determined that the change in core damage frequency result was 5.9 x 10-7. This noncited violation was therefore determined to be of very low safety significance (Green). This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because licensee personnel failed to thoroughly evaluate the multiple failures of the train B control room air conditioner compressor. [P.1(c)] (Section 4OA3.1.b) Inspection Report# : 2011002 (pdf) Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions to Address the deficiencies in the RCIC Flow Control System Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified associated with the licensees failure to take timely corrective actions to correct a condition adverse to quality associated with degradation of the Reactor Core Isolation Cooling (RCIC) flow control system. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06850. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance as it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual 0609, Significance Determination Process, Phase 1 Screening Worksheet, the inspectors determined that the finding affects the mitigating systems cornerstone because the deficiency degraded the short term heat removal capability of the RCIC system. The finding does not represent a loss of system safety function for RCIC, therefore it is determined to be of very low safety significance, or green. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources because the licensee failed to properly prioritize the work order associated with correcting the degraded condition with the RCIC flow control system [H.2(a)]. (Section 1R22) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation Following a Spurious Actuation of the Standby Service Water Pump House Ventilation Fan Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion V, involving a failure to follow procedures, which resulted in an inadequate operability evaluation. On December 5, 2010, a spurious actuation of the standby service water pump house ventilation system occurred, resulting in the pump house temperatures dropping below the design limit. The operability evaluation failed to consider the impact of the actual freezing conditions occurring at the site at that time, and operations did not secure the fan after the spurious actuation until questioned by the inspectors. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2011 00151. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance as it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety

significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to thoroughly evaluate the impact of the spurious actuation of the standby service water pump house ventilation fan [P.1 (c)]. (Section 1R15) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Required Quality Control Inspections Green. Inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. The licensee performed extensive reviews, and inspectors performed independent reviews of the licensees conclusions as well as independent sampling, to confirm that improper or missed inspections did not actually affect the operability of plant equipment. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective action program under Condition Reports CR-HQN 2009-01184 and CR-HQN-2010-0013. The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the Design Control attribute of the Mitigating Systems cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to have very low safety significance in Phase 1 of the SDP, since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this performance deficiency involved a cross-cutting aspect related to the human performance in decision-making (H.1a), because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate. [H.1(a)] (Section 4OA2) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program Green. Inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program. Specifically, the individual assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as Condition Report CR-HQN-2010-00386. Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative

assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no cross-cutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago. (Section 4OA2) Inspection Report# : 2010005 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Degraded Fire Door Barrier Protecting the Safeguards Switchgear Rooms Green. The inspectors identified a Green noncited violation of Facility Operating License Condition 2.C(41) involving the failure to ensure that fire barriers protecting safety-related areas were functional during monthly fire barrier inspections. The inspectors identified that fire door OC211, crossover door between division 1 and 2 switchgear rooms, was missing 5 screws in the divider overlap and there was a three inch crack in the door on the bottom left side. The Fire Hazards Analysis Report, Section 9A.2.4 defines fire doors as a fire barrier, and Section 9A.5.7 and 9A.5.8, Fire Area 7 and Fire Area 8, respectively, describe the electrical switchgear rooms as having 3-hour fire rated barriers. Operations initiated an hourly fire watch for the non-functional door per the technical requirements manual. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2010-05541. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety Mitigating Systems (MS) Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire confinement category. The inspectors assigned a low degradation rating because the cracks did not create an actual hole through the door. The inspectors concluded that the finding was of very low safety significance (Green) because the degraded barrier was expected to maintain nearly the same level of effectiveness and reliability had the degradation not been present, and there were no fire ignition sources or combustible materials in the area that would subject the barrier to direct flame impingement. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources, because plant personnel failed to adequately evaluate and provide proper maintenance for degrading fire doors [H.2(d)] (Section 1R05) Inspection Report# : 2010004 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Remove Foreign Material from the Control Room Air Conditioning Systems Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, for the failure to remove foreign material from the control room air conditioning oil and Freon subsystems. The pencil strainer on the compressor was found to be 90 percent clogged by foreign material. Plant personnel cleaned the pencil strainer, but placed the CRAC B system back in service without cleaning the oil and Freon subsystems which resulted in the CRAC B system becoming inoperable two weeks later. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-04839. The finding was more than minor because it was associated with the equipment performance attribute of the reactor safety Mitigating Systems (MS) Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance since it did not result in a loss of system safety function. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to appropriately address the foreign material in the control room air conditioning subsystems [P.1(d)] (Section 1R15). Inspection Report# : 2010004 (pdf)

Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Implement an Adequate Structural Monitoring Program Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B Criterion V for failure to perform required inspection of safety-related plant structures. Specifically, the inspectors found inspections that had been only partially performed and some areas that had not been documented as inspected. Subsequent walkdowns identified several deficiencies including concrete cracks and spalling, deficient coatings, rusted tanks and exposed rebar. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06871. The finding is greater than minor because it is associated with the Mitigating Systems (MS) Cornerstone attribute of protection against external events and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance since it did not represent a loss of system safety function, an actual loss of safety function of a single train for greater than its TS allowed outage time, or screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of human performance, associated with the resources component, in that the licensee failed to accurately document and manage the structural inspections [H.2(c)] (Section 40A2). Inspection Report# : 2010004 (pdf) Significance: Jun 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Risk Assessment for Emergent Work Activities

  • Green. The inspectors identified a noncited violation of 10 CFR 50.65 a(4) for failure to perform adequate risk assessments prior to flushing the reactor heat removal systems suction piping and filling and venting of the alternate decay heat removal system. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2010 02553.

This finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone, adversely affecting the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors evaluated the finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. Using Appendix K, the inspectors determined that the finding has a very low safety significance because the finding was related only to the performance of risk management actions and did not exceed the threshold for core damage probability and large early release probability. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources, because the licensee failed to provide adequate training on the implementation of the new risk management procedure [H.2(b)] (Section 1R13.b). Inspection Report# : 2010003 (pdf) Barrier Integrity Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustible Stored in the Fire Exclusion Zone Near the Independent Spent Fuel Storage Installation Green

. The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that transient combustible were not stored in the fire exclusion zone near the independent spent fuel storage installation. The inspectors performed a quarterly fire protection inspection of independent spent fuel storage installation and identified a large air conditioner with combustible material covering it located in the fire exclusion zone that was within 60 feet of the dry fuel storage pad. The inspectors determined through interviews that the material had been placed there the previous day by the maintenance department. As immediate corrective action the licensee removed the combustible material from the area. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00455. This finding was more than minor because it was associated human performance attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire prevention and administrative controls category. The inspectors assigned a low degradation rating due to the fact that the amount of combustible material in the area was minimal. The inspectors concluded that the finding was of very low safety significance (Green) due to the fact there were no fire ignition sources in the area. The cause of this finding has a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively communicate expectations regarding storage of combustible material near the dry fuel storage pad. [H.4(b)] (Section 1R05.1.b) Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Design Control for the Mitigation Monitoring System Modification Green. The inspectors reviewed a self-revealing, Green finding of EN-DC-115, Engineering Change Process, involving the failure to maintain adequate design control measures associated with the installation of the mitigation monitoring system. On November 8, 2010, a reactor coolant pressure boundary failure occurred at the skid mounted Online Noble Chemical - Mitigation Monitoring System pump inside primary containment. The positive displacement sample pump ejected the pump piston from the housing, resulting in an approximate 7 gpm leak of reactor coolant. The steam leak resulted in a reactor recirculation system flow control valve lockup (due to hydraulic power unit motor failure) and approximately 15,000 square feet of contaminated area in the primary containment structure. The licensee failed to ensure proper validation testing for the pump prior to installation. Specifically, the licensee did not ensure that the pump could withstand the operating pressures and temperatures of the system in which it was installed. The licensee removed the mitigation monitoring system from service and isolated the skid from the reactor water cleanup system. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2010-07852. The finding is more than minor because it affects the design control attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Therefore, using inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet for LOCA initiators, the inspectors concluded that the finding was of very low safety significance (Green) because the failure of the mitigation monitoring system would not have exceeded technical specifications limits for identified leakage in the reactor coolant system. This finding has a crosscutting aspect in the work practices component of the human performance area; because the licensee failed to adequately oversee the design of the mitigation monitoring system such that nuclear safety is supported. [H.4(c)] (Section 4OA3.2.b) Inspection Report# : 2011002 (pdf) Significance: Jun 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Loss of Decay Heat Removal to the Spent Fuel Pool Green. The inspectors reviewed a self-revealing noncited violation of Technical Specifications 5.4.1(a), involving a loss of decay heat removal in the spent fuel pool due to station operators failing to follow the fuel pool cooling and cleanup system operating instruction. The licensee entered this issue into the corrective action program as Condition Report CR GGN 2010 02172.

This finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using the Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding has a very low safety significance because it only represented a loss of spent fuel pool cooling that would not preclude restoration of cooling to the spent fuel pool prior to pool boiling. The cause of this finding has a crosscutting aspect in the area of human performance associated with work practices, because licensee personnel failed to use adequate self- and peer-checking techniques to remove the filter/demineralizer from service [H.4(a)] (Section 1R15.b). Inspection Report# : 2010003 (pdf) Emergency Preparedness Significance: Dec 31, 2010 Identified By: NRC Item Type: VIO Violation Failure to Have Guidelines for the Choice of Protective Actions During an Emergency Consistent with Federal Guidance Green. A cited violation of 10 CFR 50.47(b)(10) was identified for failure to develop and have in place guidelines for the choice of protective actions during an emergency that were consistent with federal guidance. Federal guidance for the choice of protective actions during an emergency is described in EPA-400-R-92-001 and states, in part, that evacuation is seldom justified when doses are less than protective action guides. The licensees automatic process that extended existing protective action recommendations with changes in wind direction without considering radiation dose was identified as a performance deficiency. This finding is more than minor because it affects the Emergency Preparedness Cornerstone objective of implementing adequate measures to protect the health and safety of the public during a radiological emergency, and is associated with the cornerstone attributes of emergency response organization performance and procedure quality. This finding was determined to be of very low safety significance because it was a failure to comply with NRC requirements, was associated with risk significant planning standard 10 CFR 50.47(b)(10), and was not a risk significant planning standard functional failure or a planning standard degraded function. The finding was not a functional failure or degraded planning standard function because appropriate protective action recommendations for the public would have been made for all areas where protective action guides were exceeded. This finding is a cited violation of 10 CFR 50.47(b)(10) because the licensee failed to restore compliance with NRC requirements in a timely manner. The finding is related to the corrective action element of the problem identification and resolution crosscutting aspect because the licensee failed to take corrective actions to address the safety issue in a timely manner [P1.d] (Section 1EP5) Inspection Report# : 2010005 (pdf) Significance: Sep 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Primary Meteorological Tower Inoperable Without Compensatory Actions in Place Green. A self-revealing non-cited violation of 10 CFR 50.47(b)(8), was identified when the Grand Gulf Nuclear Station Primary Meteorological Tower was rendered inoperable without compensatory actions from July 6 through July 27, 2010. The primary meteorological tower was declared inoperable by operations for maintenance to perform surveillance and preventative maintenance activities. The technicians did not finish the surveillance due to problems with data points exceeding allowable tolerance limits, and left the tower with the 10 and 50 meter instruments lowered to the ground. Inaccurate meteorological data continued to be displayed in the plant computer system. During the subsequent night shift, the control room supervisor inadvertently closed out the limiting condition of operations for

the primary meteorological tower being out of service prior to the tower being returned to an operable condition. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2010-05748. The finding was more than minor because it was associated with the Facilities and Equipment attribute of the Emergency Preparedness (EP) Cornerstone and adversely affected the cornerstone objective of ensuring the capability to implement adequate measures to protect public health and safety in the event of a radiological emergency. Specifically, from July 6 through July 27, 2010, key emergency response members could not have accurately performed their assigned emergency notification and dose assessment functions, with an absence of compensatory measures. In accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1 and the examples contained in section 4.8 of the same document, the inspectors determined the finding to be of very low safety significance (Green) because the performance deficiency was a failure to comply with NRC regulations, the deficiency was associated with a non-risk significant planning standard as defined in MC0609 Appendix B, and it did not represent a functional failure of the planning standard. The cause of this finding has a crosscutting aspect in the area of human performance associated with work control, because the maintenance and operations department failed to appropriately communicate and coordinate work activities on the primary meteorological tower. [H.3(b)] (Section 1R19). Inspection Report# : 2010004 (pdf) Occupational Radiation Safety Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area Green. The inspectors identified a noncited violation of Technical Specification 5.7.2, resulting from the licensees failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area. The finding was placed into the corrective action program as Condition Report CR-GGN-2011-01045, and corrective action was being evaluated. The failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area is a performance deficiency. The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to use qualified radiation protection technicians to provide job coverage in a high radiation area with dose rates in excess of 1000 mrem/hr had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. (Section 2RS01.b) Inspection Report# : 2011002 (pdf) Significance: Sep 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow the Radiological Protection Job Coverage Procedure Green. The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1.a for a procedure violation. Radiation Work Permit 20101704 covered work on Valve 1G33F253 in the reactor water cleanup room. Work on this valve was conducted over a 6-day period, May 6 through 11, 2010, and in that time, three personnel contaminations occurred. Appropriate protective clothing was not assigned by the job coverage technician and contributed to the three personnel contaminations and radioactive intake by one of the workers of 62 mrem. The failure to assign appropriate protective clothing during radiological work is a performance deficiency. The finding is greater than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and

process (exposure control), and affected the cornerstone objective, in that it resulted in an individual receiving unplanned dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work practices, because the radiation protection technician covering the job did not use risk insights or take the job site condition into consideration when assigning protective clothing for radiological work [H.3(a)] (2RSO4). Inspection Report# : 2010004 (pdf) Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : June 07, 2011

Grand Gulf 1 2Q/2011 Plant Inspection Findings Initiating Events Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Reactor Shutdown Procedure Causes Power and Level Oscillations Green. The inspectors identified a green, noncited violation of 10 CFR 50 Appendix B, Criterion V, for an inadequate shutdown procedure resulting in power and level oscillations in the reactor. The revised procedure failed to require the plant to be placed in startup feedwater level control during low power operations. In addition, the operators performed shutdown training on the old procedure. The performance deficiency was self-revealing, however the inspectors added significant value by identifying inadequate condition report classification, causal evaluation, and corrective actions. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-05140. The finding is more than minor because it was associated with the initiating events cornerstone attribute of procedure quality and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the inspectors determined that the finding impacted both the initiating events and mitigating systems cornerstones. The inspectors determined that the initiating event cornerstone best reflected the dominant risk of the finding. The finding was determined to be of very low safety significance (Green) because the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision-making, because station management failed to use conservative assumptions to demonstrate that the change to the shutdown operating procedure was safe prior to proceeding [H.1(b)]. (Section 4OA2) Inspection Report# : 2010005 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Risk Assessment for Switchyard Battery Replacement Green. The inspectors identified a Green noncited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for failure to properly assess the risk impact of maintenance on the switchyard batteries. Specifically, plant personnel evaluated the work as light switchyard work when it should have been evaluated as heavy equipment, which increases the likelihood of a loss of offsite power transient. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06668. The finding was more than minor because it was associated with the Protection Against External Factors attribute of the Initiating Event (IE) Cornerstone. Because the finding affects the licensees assessment of risk associated with performing maintenance activities, IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, directs significance determination via the use of IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Flowchart 1 of Appendix K, the significance of this finding was determined to be of very low safety significance (Green), because the calculated Incremental Core Damage Probability Deficit (2E-8) was not greater than 1.0E-6. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to use a systematic decision making process using available risk assessment guidance and did not obtain interdisciplinary input on an important risk management decision [H.1(a)] (Section 1R13). Inspection Report# : 2010004 (pdf)

Mitigating Systems Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation Door Seals Protecting Safety Related Equipment Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. Inspectors found the entrance door to the diesel generator building and the entrance door to the division 2 diesel generator in a degraded condition. The inspectors identified that the door seals did not make complete contact with the door frames all the way around as required by procedure. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-02575. The finding is more than minor because it is associated with the protection against external factors attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multiple trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis. The result was a delta-core damage frequency of 3.3E-7/yr and a delta-large early release frequency of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The inspectors determined the apparent cause of this finding was that licensee personnel were not adequately trained to perform these inspections. Therefore this finding has a cross-cutting aspect in the area of human performance associated with resources in that the licensees training of personnel was not adequate in performing inspection of the probable maximum precipitation door seals [H.2(b)](Section 1R01). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Scaffold Control Procedure Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement scaffolding control procedural requirements related to post-installation inspections and engineering safety evaluations for scaffolding constructed within 2 inches of safety-related or fire protection equipment. During plant walkdowns, inspectors identified multiple examples of the licensee not properly implementing Entergys corporate and site procedures for the control of scaffolding. The licensees immediate corrective actions included inspecting the scaffolding that had been installed, modifying or removing it where appropriate, and properly posting the scaffolds. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2011-03480, CR-GGN-2011-03601, CR-GGN-2011-03602, and CR-GGN-2011-03603. The inspectors determined that this finding is more than minor because it is associated with the external factors and equipment performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green), because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the apparent cause of this finding was lack of supervisor oversight during scaffold construction. Therefore the finding has a cross-cutting aspect in the area of human performance associated with work practices, in that the licensee did not provide effective supervisor oversight of workers constructing scaffolding to ensure these activities were performed per procedural requirements [H.4(c)](Section 1R04). Inspection Report# : 2011003 (pdf)

Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Conditions Adverse to Fire Protection Green. The inspectors identified a noncited violation of License Condition 2.C(41) for the failure to identify conditions adverse to the fire protection program. Specifically, during required inspections of the material condition of the sprinkler system, the licensee failed to identify several instances of bent or misaligned sprinkler head deflector plates and a painted sprinkler head. Corrective action included correcting bent or misaligned plates and replacing the painted sprinkler head. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-03132. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the safety concern is that the number of bent or misaligned sprinkler heat canopies and painted sprinkler heads would not provide an adequate area-wide coverage of suppression. The inspectors evaluated the significance of this finding using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The deficiency involved the Fixed Fire Protection Systems category. Using Appendix F, Enclosure , "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," the inspectors determined that the deficiency had low degradation since less than 10 percent of the heads in the affected fire area were nonfunctional, a functional head remained within 10 feet of the combustibles of concern, and the system remained nominally code compliant. This finding screened as having very low safety significance (Green) in Phase 1of Manual Chapter 0609, Appendix F. This finding has a cross-cutting aspect in the area of human performance associated with resources because the procedure used to inspect the condition of these sprinklers did not contain specific criteria for identifying unacceptable sprinkler conditions [H.2(c)](Section 1R05). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure that Safety Related Manholes were Properly Sealed to Prevent the Entry of Flammable Liquid Green. The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that manholes MH01, MH20 and MH21 were properly sealed to prevent the entry of flammable liquid. During the performance of the manhole/vault inspection, the inspectors were reviewing engineering change packages associated with solar sump pumps for MH20 and MH21. During their review, they determined that the licensee was not meeting the requirements of their license bases documents for MH20 and MH21, which contain safe shutdown cables for standby service water trains A and B. The licensees immediate corrective action included placing hazmat barricades around each manhole to prevent flammable fluids from entering the manholes. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00562. This finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 3b, Item 1 directs the inspectors to Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. However, an NRC senior reactor analyst determined that the unique nature of this performance deficiency did not lend itself to analysis by the methods provided in Appendix F. Therefore, a Phase 3 analysis was performed. Based on a bounding analysis, the analyst determined that the change in core damage frequency was approximately 1.5E-7/yr. The result was low because of the relatively short periods of time that fuel was actually being transferred, the low probability of transfer system failures, and the low likelihood that a loss of normal service water initiator would occur following a fire in the subject manholes. This noncited violation was therefore determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to initiate a condition report when the issue was identified during the development of their engineering change package, which resulted in the failure to ensure the safety related manholes were sealed in accordance with their license based documents [P.1(a)](Section 1R06).

Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Provide Adequate Procedures for High Pressure Core Spray Minimum Flow Valve Surveillance Testing Green. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for the licensees failure to provide adequate testing procedures, which resulted in the high pressure core spray minimum flow valve inadvertently stroking approximately 11 times during a surveillance test. The excessive stroking of the valve resulted in the unplanned inoperability of the high pressure core spray system because the valves feeder breaker overcurrent instantaneous trip setpoint had drifted below the manufacturers tolerance for the existing setting. As immediate corrective action, the licensee replaced the degraded breaker. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01901. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function since the high pressure core spray system would still have been functional even with the minimum flow valve potentially failing open. Additionally, it did not represent a loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with operating experience in that licensee had not incorporated operating experience from a similar event that had occurred at another Entergy site [P.2(b)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Loose Fuse Clips in Division 3 Emergency Diesel Generator Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take adequate corrective actions for a significant condition adverse to quality associated with the division 3 emergency diesel generator. While performing a maintenance effectiveness review of the diesel generators, the inspectors noted on October 17, 2009, at 9:07 p.m., the FU-7 fuse for the division 3 diesel generator was determined to have a faulty fuse clip, resulting in the inoperability of the diesel generator due to loss of power to the direct current powered fuel pumps. Then on March 18, 2011, the division 3 emergency diesel generator was again rendered inoperable due to a faulty fuse clip on the FU-8 fuse holder, which is of the same design and function as the FU-7 fuse holder in the previous occurrence. Short term corrective action included replacing the fuse holder. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01868. The finding is more than minor because it is associated with equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, 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 in the area of human performance associated with resources because the training provided to correct the initial event was not adequate to ensure proper fuse installation and verify good connection existed between the fuse and fuse holder [H.2(b)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011

Identified By: NRC Item Type: NCV NonCited Violation Failure to Assure Configuration Control of Safety Related Systems Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to review the suitability of leaving test fittings on reactor coolant system flow transmitter equalizing block drain ports instead of the design specified manifold plugs. As corrective action, the licensee replaced the test fittings with the correct drain plugs. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-04485. This finding is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability of functionality, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, associated with work practices, because the licensee failed to ensure that human error prevention techniques, such as holding pre-job briefings, self- and peer-checking, and proper documentation of activities were utilized such that work activities were performed safely and personnel did not proceed in the face of uncertainty or unexpected circumstances Specifically, the licensee failed to review the suitability of installing test and brass fittings on pressure, differential pressure and flow transmitter block valve drain ports instead of the design specified manifold plugs. [H.4(a)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow a Procedure Resulting in the Inoperability of the Reactor Core Isolation Cooling System Primary Containment Isolation Valve Green. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, for failure to follow a procedure resulting in the inoperability of the reactor core isolation cooling system primary containment isolation valve. This occurred while the licensee was performing surveillance on the reactor core isolation cooling system and incorrectly attached a jumper to the wrong terminal point resulting in blowing a fuse that caused a loss of control power to the reactor core isolation cooling primary containment isolation valve 1E51-F031. As immediate corrective action, the licensee removed the jumper and replaced the control power fuse. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01932. The finding is more than minor since it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. In addition, this finding had a human performance cross-cutting aspect associated with work practices in that the licensee did not use the proper human performance techniques of self-checking to prevent the loss of control power to a primary containment isolation valve [H.4(a)](Section 1R22). Inspection Report# : 2011003 (pdf) Significance: SL-IV Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update Available Low Pressure Coolant Injection Loops in the Updated Final Safety Analysis Report SLIV. Inspectors identified a noncited violation of 10 CFR 50.71(e)(4), which requires the final safety analysis report be updated, at intervals not exceeding 24 months, to reflect changes made in the facility or procedures described in the final safety analysis report. Licensee personnel failed to update the original revision of the final safety analysis report to reflect the actual number of low pressure coolant injection loops available for automatic initiation during shutdown

cooling operations in Mode 3. The licensee plans to update the final safety analysis report at the next scheduled revision. This finding was entered into the licensees corrective action program as condition report CR-GGN-2011-01631. The failure of licensing personnel to update the final safety analysis report to reflect the available low pressure coolant injection loops for automatic initiation during shutdown cooling operations in Mode 3 was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy, dated September 30, 2010, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate Maintenance Effectiveness of Train B Control Room Air Conditioner Green. The inspectors identified a noncited violation of 10 CFR Part 50.65(a)(2) for the licensees failure to demonstrate that the performance of the train B control room air conditioner was being effectively controlled through the performance of appropriate preventive maintenance. Engineering did not properly evaluate maintenance rule functional failures resulting in the system remaining in an a(2) status instead of an a(1) status. As corrective action, the train B control room air conditioner was moved into an a(1) status. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2011-01623. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. As corrective action, the train B control room air conditioner was moved into an (a)(1) status. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because licensee personnel failed to make appropriate safety-significant or risk-significant decisions to address the multiple failures of the train B control room air conditioner compressor. [H.1(a)] (Section 1R12.b.2) Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Control Room Air Conditioner Compressor Tripping Due to Low Oil Pressure Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after the licensee failed to determine the cause and prevent recurrence of a significant condition adverse to quality associated with the train B control room air conditioner compressor tripping due to low oil pressure. Specifically, on December 13, 2010, the train B control room air conditioner compressor tripped on low oil pressure after the licensee had performed a root cause analysis to identify the cause and prevent recurrence of a similar compressor trip on October 14, 2010. As immediate corrective action, the licensee installed an inline suction filter. No additional failures have occurred since its installation. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2010-07315. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, the inspectors determined that a Phase 2 analysis was required because the finding represented a loss of system safety function. The plant-specific risk informed notebook does not include the evaluation of risk caused by the loss of cooling to the main control room. Therefore, the senior reactor analyst conducted a Phase 3 analysis. Based on the bounding analysis, the analyst determined that the change in core damage frequency result was 5.9 x 10-7. This

noncited violation was therefore determined to be of very low safety significance (Green). This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because licensee personnel failed to thoroughly evaluate the multiple failures of the train B control room air conditioner compressor. [P.1(c)] (Section 4OA3.1.b) Inspection Report# : 2011002 (pdf) Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Untimely Corrective Actions to Address the deficiencies in the RCIC Flow Control System Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified associated with the licensees failure to take timely corrective actions to correct a condition adverse to quality associated with degradation of the Reactor Core Isolation Cooling (RCIC) flow control system. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06850. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance as it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual 0609, Significance Determination Process, Phase 1 Screening Worksheet, the inspectors determined that the finding affects the mitigating systems cornerstone because the deficiency degraded the short term heat removal capability of the RCIC system. The finding does not represent a loss of system safety function for RCIC, therefore it is determined to be of very low safety significance, or green. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources because the licensee failed to properly prioritize the work order associated with correcting the degraded condition with the RCIC flow control system [H.2(a)]. (Section 1R22) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation Following a Spurious Actuation of the Standby Service Water Pump House Ventilation Fan Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion V, involving a failure to follow procedures, which resulted in an inadequate operability evaluation. On December 5, 2010, a spurious actuation of the standby service water pump house ventilation system occurred, resulting in the pump house temperatures dropping below the design limit. The operability evaluation failed to consider the impact of the actual freezing conditions occurring at the site at that time, and operations did not secure the fan after the spurious actuation until questioned by the inspectors. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2011 00151. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance as it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to thoroughly evaluate the impact of the spurious actuation of the standby service water pump house ventilation fan [P.1 (c)]. (Section 1R15) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC

Item Type: NCV NonCited Violation Failure to Perform Required Quality Control Inspections Green. Inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. The licensee performed extensive reviews, and inspectors performed independent reviews of the licensees conclusions as well as independent sampling, to confirm that improper or missed inspections did not actually affect the operability of plant equipment. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective action program under Condition Reports CR-HQN 2009-01184 and CR-HQN-2010-0013. The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the Design Control attribute of the Mitigating Systems cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to have very low safety significance in Phase 1 of the SDP, since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this performance deficiency involved a cross-cutting aspect related to the human performance in decision-making (H.1a), because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate. [H.1(a)] (Section 4OA2) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program Green. Inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program. Specifically, the individual assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as Condition Report CR-HQN-2010-00386. Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no cross-cutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago. (Section 4OA2) Inspection Report# : 2010005 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation

Degraded Fire Door Barrier Protecting the Safeguards Switchgear Rooms Green. The inspectors identified a Green noncited violation of Facility Operating License Condition 2.C(41) involving the failure to ensure that fire barriers protecting safety-related areas were functional during monthly fire barrier inspections. The inspectors identified that fire door OC211, crossover door between division 1 and 2 switchgear rooms, was missing 5 screws in the divider overlap and there was a three inch crack in the door on the bottom left side. The Fire Hazards Analysis Report, Section 9A.2.4 defines fire doors as a fire barrier, and Section 9A.5.7 and 9A.5.8, Fire Area 7 and Fire Area 8, respectively, describe the electrical switchgear rooms as having 3-hour fire rated barriers. Operations initiated an hourly fire watch for the non-functional door per the technical requirements manual. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2010-05541. The finding was more than minor since it was associated with the protection against external factors attribute of the reactor safety Mitigating Systems (MS) Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire confinement category. The inspectors assigned a low degradation rating because the cracks did not create an actual hole through the door. The inspectors concluded that the finding was of very low safety significance (Green) because the degraded barrier was expected to maintain nearly the same level of effectiveness and reliability had the degradation not been present, and there were no fire ignition sources or combustible materials in the area that would subject the barrier to direct flame impingement. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources, because plant personnel failed to adequately evaluate and provide proper maintenance for degrading fire doors [H.2(d)] (Section 1R05) Inspection Report# : 2010004 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Remove Foreign Material from the Control Room Air Conditioning Systems Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, for the failure to remove foreign material from the control room air conditioning oil and Freon subsystems. The pencil strainer on the compressor was found to be 90 percent clogged by foreign material. Plant personnel cleaned the pencil strainer, but placed the CRAC B system back in service without cleaning the oil and Freon subsystems which resulted in the CRAC B system becoming inoperable two weeks later. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-04839. The finding was more than minor because it was associated with the equipment performance attribute of the reactor safety Mitigating Systems (MS) Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance since it did not result in a loss of system safety function. The cause of this finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to appropriately address the foreign material in the control room air conditioning subsystems [P.1(d)] (Section 1R15). Inspection Report# : 2010004 (pdf) Significance: Sep 27, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Implement an Adequate Structural Monitoring Program Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B Criterion V for failure to perform required inspection of safety-related plant structures. Specifically, the inspectors found inspections that had been only partially performed and some areas that had not been documented as inspected. Subsequent walkdowns identified several deficiencies including concrete cracks and spalling, deficient coatings, rusted tanks and exposed rebar. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06871. The finding is greater than minor because it is associated with the Mitigating Systems (MS) Cornerstone attribute of protection against external events and affects the cornerstone objective to ensure the availability, reliability, and

capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, the finding is determined to have very low safety significance since it did not represent a loss of system safety function, an actual loss of safety function of a single train for greater than its TS allowed outage time, or screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of human performance, associated with the resources component, in that the licensee failed to accurately document and manage the structural inspections [H.2(c)] (Section 40A2). Inspection Report# : 2010004 (pdf) Barrier Integrity Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustible Stored in the Fire Exclusion Zone Near the Independent Spent Fuel Storage Installation Green . The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that transient combustible were not stored in the fire exclusion zone near the independent spent fuel storage installation. The inspectors performed a quarterly fire protection inspection of independent spent fuel storage installation and identified a large air conditioner with combustible material covering it located in the fire exclusion zone that was within 60 feet of the dry fuel storage pad. The inspectors determined through interviews that the material had been placed there the previous day by the maintenance department. As immediate corrective action the licensee removed the combustible material from the area. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00455. This finding was more than minor because it was associated human performance attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire prevention and administrative controls category. The inspectors assigned a low degradation rating due to the fact that the amount of combustible material in the area was minimal. The inspectors concluded that the finding was of very low safety significance (Green) due to the fact there were no fire ignition sources in the area. The cause of this finding has a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively communicate expectations regarding storage of combustible material near the dry fuel storage pad. [H.4(b)] (Section 1R05.1.b) Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Design Control for the Mitigation Monitoring System Modification Green. The inspectors reviewed a self-revealing, Green finding of EN-DC-115, Engineering Change Process, involving the failure to maintain adequate design control measures associated with the installation of the mitigation monitoring system. On November 8, 2010, a reactor coolant pressure boundary failure occurred at the skid mounted Online Noble Chemical - Mitigation Monitoring System pump inside primary containment. The positive displacement sample pump ejected the pump piston from the housing, resulting in an approximate 7 gpm leak of reactor coolant. The steam leak resulted in a reactor recirculation system flow control valve lockup (due to hydraulic power unit motor failure) and approximately 15,000 square feet of contaminated area in the primary containment structure. The licensee failed to ensure proper validation testing for the pump prior to installation. Specifically, the licensee did not ensure that the pump could withstand the operating pressures and temperatures of the system in which it was installed. The licensee removed the mitigation monitoring system from service and isolated the skid from the reactor water cleanup system. This finding was entered into the licensees corrective action program as

Condition Report CR-GGN-2010-07852. The finding is more than minor because it affects the design control attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Therefore, using inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet for LOCA initiators, the inspectors concluded that the finding was of very low safety significance (Green) because the failure of the mitigation monitoring system would not have exceeded technical specifications limits for identified leakage in the reactor coolant system. This finding has a crosscutting aspect in the work practices component of the human performance area; because the licensee failed to adequately oversee the design of the mitigation monitoring system such that nuclear safety is supported. [H.4(c)] (Section 4OA3.2.b) Inspection Report# : 2011002 (pdf) Emergency Preparedness Significance: Dec 31, 2010 Identified By: NRC Item Type: VIO Violation Failure to Have Guidelines for the Choice of Protective Actions During an Emergency Consistent with Federal Guidance Green. A cited violation of 10 CFR 50.47(b)(10) was identified for failure to develop and have in place guidelines for the choice of protective actions during an emergency that were consistent with federal guidance. Federal guidance for the choice of protective actions during an emergency is described in EPA-400-R-92-001 and states, in part, that evacuation is seldom justified when doses are less than protective action guides. The licensees automatic process that extended existing protective action recommendations with changes in wind direction without considering radiation dose was identified as a performance deficiency. This finding is more than minor because it affects the Emergency Preparedness Cornerstone objective of implementing adequate measures to protect the health and safety of the public during a radiological emergency, and is associated with the cornerstone attributes of emergency response organization performance and procedure quality. This finding was determined to be of very low safety significance because it was a failure to comply with NRC requirements, was associated with risk significant planning standard 10 CFR 50.47(b)(10), and was not a risk significant planning standard functional failure or a planning standard degraded function. The finding was not a functional failure or degraded planning standard function because appropriate protective action recommendations for the public would have been made for all areas where protective action guides were exceeded. This finding is a cited violation of 10 CFR 50.47(b)(10) because the licensee failed to restore compliance with NRC requirements in a timely manner. The finding is related to the corrective action element of the problem identification and resolution crosscutting aspect because the licensee failed to take corrective actions to address the safety issue in a timely manner [P1.d] (Section 1EP5) Inspection Report# : 2010005 (pdf) Significance: Sep 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Primary Meteorological Tower Inoperable Without Compensatory Actions in Place Green. A self-revealing non-cited violation of 10 CFR 50.47(b)(8), was identified when the Grand Gulf Nuclear Station Primary Meteorological Tower was rendered inoperable without compensatory actions from July 6 through July 27, 2010. The primary meteorological tower was declared inoperable by operations for maintenance to perform surveillance and preventative maintenance activities. The technicians did not finish the surveillance due to problems with data points exceeding allowable tolerance limits, and left the tower with the 10 and 50 meter instruments lowered to the ground. Inaccurate meteorological data continued to be displayed in the plant computer system. During the subsequent night shift, the control room supervisor inadvertently closed out the limiting condition of operations for the primary meteorological tower being out of service prior to the tower being returned to an operable condition. The

licensee entered this issue into their corrective action program as Condition Report CR GGN 2010-05748. The finding was more than minor because it was associated with the Facilities and Equipment attribute of the Emergency Preparedness (EP) Cornerstone and adversely affected the cornerstone objective of ensuring the capability to implement adequate measures to protect public health and safety in the event of a radiological emergency. Specifically, from July 6 through July 27, 2010, key emergency response members could not have accurately performed their assigned emergency notification and dose assessment functions, with an absence of compensatory measures. In accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1 and the examples contained in section 4.8 of the same document, the inspectors determined the finding to be of very low safety significance (Green) because the performance deficiency was a failure to comply with NRC regulations, the deficiency was associated with a non-risk significant planning standard as defined in MC0609 Appendix B, and it did not represent a functional failure of the planning standard. The cause of this finding has a crosscutting aspect in the area of human performance associated with work control, because the maintenance and operations department failed to appropriately communicate and coordinate work activities on the primary meteorological tower. [H.3(b)] (Section 1R19). Inspection Report# : 2010004 (pdf) Occupational Radiation Safety Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area Green. The inspectors identified a noncited violation of Technical Specification 5.7.2, resulting from the licensees failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area. The finding was placed into the corrective action program as Condition Report CR-GGN-2011-01045, and corrective action was being evaluated. The failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area is a performance deficiency. The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to use qualified radiation protection technicians to provide job coverage in a high radiation area with dose rates in excess of 1000 mrem/hr had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. (Section 2RS01.b) Inspection Report# : 2011002 (pdf) Significance: Sep 27, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow the Radiological Protection Job Coverage Procedure Green. The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1.a for a procedure violation. Radiation Work Permit 20101704 covered work on Valve 1G33F253 in the reactor water cleanup room. Work on this valve was conducted over a 6-day period, May 6 through 11, 2010, and in that time, three personnel contaminations occurred. Appropriate protective clothing was not assigned by the job coverage technician and contributed to the three personnel contaminations and radioactive intake by one of the workers of 62 mrem. The failure to assign appropriate protective clothing during radiological work is a performance deficiency. The finding is greater than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process (exposure control), and affected the cornerstone objective, in that it resulted in an individual receiving

unplanned dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding has a human performance crosscutting aspect associated with work practices, because the radiation protection technician covering the job did not use risk insights or take the job site condition into consideration when assigning protective clothing for radiological work [H.3(a)] (2RSO4). Inspection Report# : 2010004 (pdf) Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : October 14, 2011

Grand Gulf 1 3Q/2011 Plant Inspection Findings Initiating Events Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Reactor Shutdown Procedure Causes Power and Level Oscillations Green. The inspectors identified a green, noncited violation of 10 CFR 50 Appendix B, Criterion V, for an inadequate shutdown procedure resulting in power and level oscillations in the reactor. The revised procedure failed to require the plant to be placed in startup feedwater level control during low power operations. In addition, the operators performed shutdown training on the old procedure. The performance deficiency was self-revealing, however the inspectors added significant value by identifying inadequate condition report classification, causal evaluation, and corrective actions. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-05140. The finding is more than minor because it was associated with the initiating events cornerstone attribute of procedure quality and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Screening Worksheet, the inspectors determined that the finding impacted both the initiating events and mitigating systems cornerstones. The inspectors determined that the initiating event cornerstone best reflected the dominant risk of the finding. The finding was determined to be of very low safety significance (Green) because the transient initiator did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. The cause of this finding has a crosscutting aspect in the area of human performance associated with decision-making, because station management failed to use conservative assumptions to demonstrate that the change to the shutdown operating procedure was safe prior to proceeding [H.1(b)]. (Section 4OA2) Inspection Report# : 2010005 (pdf) Mitigating Systems Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance for Components Used in Critical Applications Green. The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1.a for the licensees failure to follow a procedure that required them to evaluate components of critical systems in order to establish a preventive maintenance strategy, which resulted in unscheduled unavailability of safety-related systems and associated unscheduled entries into 72-hour shutdown Technical Specification action statements. The inspectors noted the following two examples dealing with failures of safety related equipment, which resulted in entering into shutdown limiting condition of operation. On June 2, 2011, Grand Gulf Nuclear Station experienced a failure of a relay in the standby service water B pump house ventilation system, which rendered the standby service water B system inoperable. The immediate corrective actions were to replace the relays and to restore the ventilation system. On June 22, 2011, the station experienced a failure of a motor contactor coil on breaker 52-154128, which caused the engineered safety feature electrical switchgear room cooler fan coil unit 1T46B003A not to run. The maintenance personnel determined the failure was due to a burnt motor contactor coil. The immediate corrective action was to replace the contactor coil and restore the room cooler. In both cases, the failed equipment was original plant equipment and preventive maintenance measures had not been established. The licensee entered these issues into the corrective action program as Condition Reports CR-GGN-2011-3730 and CR-GGN-2011-4313.

The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function of a single train for more than its technical specifications allowed outage time. This issue is a latent issue associated with original plant equipment and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 4OA2). Inspection Report# : 2011004 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Provide An Adequate Alternative Shutdown Procedure

  • Green. The team identified a noncited violation of License Condition 2.C(41), Fire Protection Program, for failing to ensure that the postfire safe shutdown procedure for fires requiring control room evacuation could be performed within the critical times required by the approved fire protection program. Specifically, two crews of operators simulating performance of Procedure 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Revision 036, did not give priority to the required safe shutdown components which are protected against fire damage and did not complete the equipment alignments within the times required by the thermal-hydraulic analysis. The team confirmed at the end of each walkdown that the operators involved did not know what the credited shutdown equipment was for a postfire safe shutdown or the critical time limits to be met. The team also confirmed that the licensee had not performed timed walkdowns to validate that the procedure would complete the required actions for postfire safe shutdown within the times required by the thermal-hydraulic analysis. The licensee entered this into their corrective action program as CR GGN 2011 02721, implemented compensatory measures to focus the operators priority on the required safe shutdown components and implemented a procedure revision.

The failure to provide an adequate procedure to implement the requirements of the approved fire protection program for a fire in the control room is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. The scenario impacted operators being ready to emergency depressurize the reactor and reflood using a residual heat removal pump. Because a bounding change to core damage frequency was 4.13 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding is of very low safety significance (Green). The finding did not have a crosscutting aspect since the primary cause did not fit any crosscutting aspects. (Section 1R5.5.b.1) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Assure Equipment Required For Postfire Safe Shutdown Was Protected Against Fire Damage Green. The team identified a noncited violation of License Condition 2.C(41), Fire Protection Program, for failing to assure that equipment relied upon for safe shutdown following a fire in the control room was protected against fire damage. Specifically, Procedure 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Revision 036, relied on the automatic operation of and indications from the load shedding and sequencing system. The team identified that this system was not isolated from potential damage due to a fire in the control room and the procedure did not adequately address the potential that fire damage to the system could effect the postfire safe shutdown capability by spuriously starting or stopping electric loads. The licensee entered this into their corrective action program as CR GGN 2011 02721.

The failure to assure that equipment required to successfully implement the safe shutdown procedure for a fire in the control room was protected against fire damage is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 1.97 x 10-8, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding did not have a crosscutting aspect since it was not indicative of current performance, in that the licensee had established the current procedure more than three years prior to this finding. (Section 1R5.5.b.2) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions to Protect Safe Shutdown Equipment From Fire Damage Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to take timely corrective action to modify the control circuits for 33 motor operated valves that are relied upon during safe shutdown due to fire. Noncited violation NCV 05000416/2008006-04, Failure to Ensure That Damage to Motor-Operated Valve Circuits Would Not Prevent Safe Shutdown, documented the licensees inadequate review of Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During Control Room Fire. The licensee failed to develop modification packages such that motor operated valve control circuit modifications could be implemented during the fall 2010 refueling outage. As a result, 33 motor operated valves associated with safe shutdown equipment continue to remain susceptible to potential damage during spurious operation due to circuit hot shorts. The licensee has maintained a fire watch as a compensatory measure. The licensee entered this into their corrective action program as CR GGN 2011 02779. The failure to take timely corrective actions to address the potential for fire induced hot shorts to impact the ability to safely shutdown the plant following a fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 9.58 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding had a crosscutting aspect in the area of Human Performance associated with Decision Making, because the licensee failed to demonstrate that nuclear safety is an overriding priority. Specifically, the licensee did not promptly initiate control circuit reviews and implement modifications required for corrective actions after the licensees inadequate evaluation of Information Notice 92-18 was identified in the 2008 violation. [H.1(a)] (Section 1R5.6) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions To Assure Postfire Safe Shutdown Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for inadequate corrective actions to address the potential for fire induced hot shorts to impact the ability to trip a control rod group as described in Information Notice 2007-07. The licensees evaluation of Information Notice 2007-07 stated in part, provisions have been included in 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, to trip the proper reactor protective system breakers to ensure that the reactor scram occurs. The team noted that Procedure 05-1-02-II-1 contained a conditional statement for the operator to determine if opening the reactor protective system breakers is required. The procedure did not provide assurance that all control

rod groups insert since the control room indications to be utilized by the operator were not identified and confirmed to be reliable during fires requiring control room evacuation. The licensee entered this finding into its corrective action program under CR-GGN-2011-02780, implemented compensatory measures to ensure the operators de-energized the reactor protection system, and implemented a procedure change. The failure to take adequate corrective actions to address the potential for fire induced hot shorts to impact the ability to safely shutdown the plant following a fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 9.58 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding did not have a crosscutting aspect since it was not indicative of current performance. The licensee had incorrectly assessed the applicability of Information Notice 2007-07 more than three years prior to this finding. (Section 4OA2.b) Inspection Report# : 2011007 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation Door Seals Protecting Safety Related Equipment Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. Inspectors found the entrance door to the diesel generator building and the entrance door to the division 2 diesel generator in a degraded condition. The inspectors identified that the door seals did not make complete contact with the door frames all the way around as required by procedure. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-02575. The finding is more than minor because it is associated with the protection against external factors attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multiple trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis. The result was a delta-core damage frequency of 3.3E-7/yr and a delta-large early release frequency of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The inspectors determined the apparent cause of this finding was that licensee personnel were not adequately trained to perform these inspections. Therefore this finding has a cross-cutting aspect in the area of human performance associated with resources in that the licensees training of personnel was not adequate in performing inspection of the probable maximum precipitation door seals [H.2(b)](Section 1R01). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Scaffold Control Procedure Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement scaffolding control procedural requirements related to post-installation inspections and engineering safety evaluations for scaffolding constructed within 2 inches of safety-related or fire protection equipment. During plant walkdowns, inspectors identified multiple examples of the licensee not properly implementing Entergys corporate and site procedures for the control of scaffolding. The licensees immediate corrective actions included inspecting the scaffolding that had been installed, modifying or removing it where appropriate, and properly posting the scaffolds. This issue was entered into the licensees corrective

action program as Condition Reports CR-GGN-2011-03480, CR-GGN-2011-03601, CR-GGN-2011-03602, and CR-GGN-2011-03603. The inspectors determined that this finding is more than minor because it is associated with the external factors and equipment performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green), because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the apparent cause of this finding was lack of supervisor oversight during scaffold construction. Therefore the finding has a cross-cutting aspect in the area of human performance associated with work practices, in that the licensee did not provide effective supervisor oversight of workers constructing scaffolding to ensure these activities were performed per procedural requirements [H.4(c)](Section 1R04). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Conditions Adverse to Fire Protection Green. The inspectors identified a noncited violation of License Condition 2.C(41) for the failure to identify conditions adverse to the fire protection program. Specifically, during required inspections of the material condition of the sprinkler system, the licensee failed to identify several instances of bent or misaligned sprinkler head deflector plates and a painted sprinkler head. Corrective action included correcting bent or misaligned plates and replacing the painted sprinkler head. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-03132. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the safety concern is that the number of bent or misaligned sprinkler heat canopies and painted sprinkler heads would not provide an adequate area-wide coverage of suppression. The inspectors evaluated the significance of this finding using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The deficiency involved the Fixed Fire Protection Systems category. Using Appendix F, Enclosure , "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," the inspectors determined that the deficiency had low degradation since less than 10 percent of the heads in the affected fire area were nonfunctional, a functional head remained within 10 feet of the combustibles of concern, and the system remained nominally code compliant. This finding screened as having very low safety significance (Green) in Phase 1of Manual Chapter 0609, Appendix F. This finding has a cross-cutting aspect in the area of human performance associated with resources because the procedure used to inspect the condition of these sprinklers did not contain specific criteria for identifying unacceptable sprinkler conditions [H.2(c)](Section 1R05). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure that Safety Related Manholes were Properly Sealed to Prevent the Entry of Flammable Liquid Green. The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that manholes MH01, MH20 and MH21 were properly sealed to prevent the entry of flammable liquid. During the performance of the manhole/vault inspection, the inspectors were reviewing engineering change packages associated with solar sump pumps for MH20 and MH21. During their review, they determined that the licensee was not meeting the requirements of their license bases documents for MH20 and MH21, which contain safe shutdown cables for standby service water trains A and B. The licensees immediate corrective action included placing hazmat barricades around each manhole to prevent flammable fluids from entering the manholes. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00562.

This finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 3b, Item 1 directs the inspectors to Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. However, an NRC senior reactor analyst determined that the unique nature of this performance deficiency did not lend itself to analysis by the methods provided in Appendix F. Therefore, a Phase 3 analysis was performed. Based on a bounding analysis, the analyst determined that the change in core damage frequency was approximately 1.5E-7/yr. The result was low because of the relatively short periods of time that fuel was actually being transferred, the low probability of transfer system failures, and the low likelihood that a loss of normal service water initiator would occur following a fire in the subject manholes. This noncited violation was therefore determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to initiate a condition report when the issue was identified during the development of their engineering change package, which resulted in the failure to ensure the safety related manholes were sealed in accordance with their license based documents [P.1(a)](Section 1R06). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Provide Adequate Procedures for High Pressure Core Spray Minimum Flow Valve Surveillance Testing Green. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for the licensees failure to provide adequate testing procedures, which resulted in the high pressure core spray minimum flow valve inadvertently stroking approximately 11 times during a surveillance test. The excessive stroking of the valve resulted in the unplanned inoperability of the high pressure core spray system because the valves feeder breaker overcurrent instantaneous trip setpoint had drifted below the manufacturers tolerance for the existing setting. As immediate corrective action, the licensee replaced the degraded breaker. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01901. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function since the high pressure core spray system would still have been functional even with the minimum flow valve potentially failing open. Additionally, it did not represent a loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with operating experience in that licensee had not incorporated operating experience from a similar event that had occurred at another Entergy site [P.2(b)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Loose Fuse Clips in Division 3 Emergency Diesel Generator Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take adequate corrective actions for a significant condition adverse to quality associated with the division 3 emergency diesel generator. While performing a maintenance effectiveness review of the diesel generators, the inspectors noted on October 17, 2009, at 9:07 p.m., the FU-7 fuse for the division 3 diesel generator was determined to have a faulty fuse clip, resulting in the inoperability of the diesel generator due to loss of power to the direct current powered fuel pumps. Then on March 18, 2011, the division 3 emergency diesel generator was again rendered inoperable due to a faulty fuse clip on the FU-8 fuse holder, which is of the same design and function as the FU-7 fuse holder in the previous occurrence. Short term corrective action included replacing the fuse holder. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-

2011-01868. The finding is more than minor because it is associated with equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, 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 in the area of human performance associated with resources because the training provided to correct the initial event was not adequate to ensure proper fuse installation and verify good connection existed between the fuse and fuse holder [H.2(b)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Assure Configuration Control of Safety Related Systems Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to review the suitability of leaving test fittings on reactor coolant system flow transmitter equalizing block drain ports instead of the design specified manifold plugs. As corrective action, the licensee replaced the test fittings with the correct drain plugs. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-04485. This finding is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability of functionality, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, associated with work practices, because the licensee failed to ensure that human error prevention techniques, such as holding pre-job briefings, self- and peer-checking, and proper documentation of activities were utilized such that work activities were performed safely and personnel did not proceed in the face of uncertainty or unexpected circumstances Specifically, the licensee failed to review the suitability of installing test and brass fittings on pressure, differential pressure and flow transmitter block valve drain ports instead of the design specified manifold plugs. [H.4(a)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow a Procedure Resulting in the Inoperability of the Reactor Core Isolation Cooling System Primary Containment Isolation Valve Green. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, for failure to follow a procedure resulting in the inoperability of the reactor core isolation cooling system primary containment isolation valve. This occurred while the licensee was performing surveillance on the reactor core isolation cooling system and incorrectly attached a jumper to the wrong terminal point resulting in blowing a fuse that caused a loss of control power to the reactor core isolation cooling primary containment isolation valve 1E51-F031. As immediate corrective action, the licensee removed the jumper and replaced the control power fuse. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01932. The finding is more than minor since it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a

system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. In addition, this finding had a human performance cross-cutting aspect associated with work practices in that the licensee did not use the proper human performance techniques of self-checking to prevent the loss of control power to a primary containment isolation valve [H.4(a)](Section 1R22). Inspection Report# : 2011003 (pdf) Significance: SL-IV Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update Available Low Pressure Coolant Injection Loops in the Updated Final Safety Analysis Report SLIV. Inspectors identified a noncited violation of 10 CFR 50.71(e)(4), which requires the final safety analysis report be updated, at intervals not exceeding 24 months, to reflect changes made in the facility or procedures described in the final safety analysis report. Licensee personnel failed to update the original revision of the final safety analysis report to reflect the actual number of low pressure coolant injection loops available for automatic initiation during shutdown cooling operations in Mode 3. The licensee plans to update the final safety analysis report at the next scheduled revision. This finding was entered into the licensees corrective action program as condition report CR-GGN-2011-01631. The failure of licensing personnel to update the final safety analysis report to reflect the available low pressure coolant injection loops for automatic initiation during shutdown cooling operations in Mode 3 was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy, dated September 30, 2010, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate Maintenance Effectiveness of Train B Control Room Air Conditioner Green. The inspectors identified a noncited violation of 10 CFR Part 50.65(a)(2) for the licensees failure to demonstrate that the performance of the train B control room air conditioner was being effectively controlled through the performance of appropriate preventive maintenance. Engineering did not properly evaluate maintenance rule functional failures resulting in the system remaining in an a(2) status instead of an a(1) status. As corrective action, the train B control room air conditioner was moved into an a(1) status. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2011-01623. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. As corrective action, the train B control room air conditioner was moved into an (a)(1) status. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because licensee personnel failed to make appropriate safety-significant or risk-significant decisions to address the multiple failures of the train B control room air conditioner compressor. [H.1(a)] (Section 1R12.b.2) Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Control Room Air Conditioner Compressor Tripping Due to Low Oil Pressure

Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after the licensee failed to determine the cause and prevent recurrence of a significant condition adverse to quality associated with the train B control room air conditioner compressor tripping due to low oil pressure. Specifically, on December 13, 2010, the train B control room air conditioner compressor tripped on low oil pressure after the licensee had performed a root cause analysis to identify the cause and prevent recurrence of a similar compressor trip on October 14, 2010. As immediate corrective action, the licensee installed an inline suction filter. No additional failures have occurred since its installation. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2010-07315. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, the inspectors determined that a Phase 2 analysis was required because the finding represented a loss of system safety function. The plant-specific risk informed notebook does not include the evaluation of risk caused by the loss of cooling to the main control room. Therefore, the senior reactor analyst conducted a Phase 3 analysis. Based on the bounding analysis, the analyst determined that the change in core damage frequency result was 5.9 x 10-7. This noncited violation was therefore determined to be of very low safety significance (Green). This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because licensee personnel failed to thoroughly evaluate the multiple failures of the train B control room air conditioner compressor. [P.1(c)] (Section 4OA3.1.b) Inspection Report# : 2011002 (pdf) Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Untimely Corrective Actions to Address the deficiencies in the RCIC Flow Control System Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified associated with the licensees failure to take timely corrective actions to correct a condition adverse to quality associated with degradation of the Reactor Core Isolation Cooling (RCIC) flow control system. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06850. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance as it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual 0609, Significance Determination Process, Phase 1 Screening Worksheet, the inspectors determined that the finding affects the mitigating systems cornerstone because the deficiency degraded the short term heat removal capability of the RCIC system. The finding does not represent a loss of system safety function for RCIC, therefore it is determined to be of very low safety significance, or green. The cause of this finding has a crosscutting aspect in the area of human performance associated with resources because the licensee failed to properly prioritize the work order associated with correcting the degraded condition with the RCIC flow control system [H.2(a)]. (Section 1R22) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Operability Evaluation Following a Spurious Actuation of the Standby Service Water Pump House Ventilation Fan Green. The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion V, involving a failure to follow procedures, which resulted in an inadequate operability evaluation. On December 5, 2010, a spurious actuation of the standby service water pump house ventilation system occurred, resulting in the pump house temperatures dropping below the design limit. The operability evaluation failed to consider the impact of the actual freezing conditions occurring at the site at that time, and operations did not secure the fan after the spurious actuation until questioned by the inspectors. The licensee entered this issue into their corrective action program as Condition Report CR GGN 2011 00151.

This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance as it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, this finding was of very low safety significance since it did not result in a loss of operability, nor did it screen as potentially risk significant due to a seismic, flooding, or severe weather-initiating event. The cause of this finding had a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to thoroughly evaluate the impact of the spurious actuation of the standby service water pump house ventilation fan [P.1 (c)]. (Section 1R15) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Required Quality Control Inspections Green. Inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. The licensee performed extensive reviews, and inspectors performed independent reviews of the licensees conclusions as well as independent sampling, to confirm that improper or missed inspections did not actually affect the operability of plant equipment. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective action program under Condition Reports CR-HQN 2009-01184 and CR-HQN-2010-0013. The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the Design Control attribute of the Mitigating Systems cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to have very low safety significance in Phase 1 of the SDP, since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this performance deficiency involved a cross-cutting aspect related to the human performance in decision-making (H.1a), because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate. [H.1(a)] (Section 4OA2) Inspection Report# : 2010005 (pdf) Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program Green. Inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program. Specifically, the individual assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as Condition Report CR-HQN-2010-00386. Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance

deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no cross-cutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago. (Section 4OA2) Inspection Report# : 2010005 (pdf) Barrier Integrity Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Ensure Correct Fuses were Installed in the Hydrogen Igniter Control Circuits Green. The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to install the correct amperage fuses in the hydrogen igniter control circuit. On August 4, 2011, the inspectors were performing an operability review of a condition report dealing with the division 1 hydrogen igniters. The licensee had determined that half the division 1 hydrogen igniters would not energize, and in their investigation, they determined that the loss of power to the hydrogen igniters was due to a blown fuse. The licensee also determined that the blown fuse was 0.3 amps and should have been 0.8 amps per plant drawings. The licensee performed an operability determination for the as found condition and determined that the circuit required 0.193 amps to power the circuit, which included the light bulbs. The inspectors reviewed the operability determination and the calculations and determined that the licensees conclusions were reasonable. The licensee immediate corrective action was to replace the incorrect fuses one division at a time with the correct size 0.8 amp fuses and restore the hydrogen igniters to operable status. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-005388. This finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone, and it adversely affected the cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that Appendix H, Containment Integrity Significance Determination Process, was required. Inspectors determined that this was a type B finding and, using section 6.0, determined that the finding was of very low safety significance (Green) because during their review, the inspectors noted that the hydrogen igniters had maintain functionality over the life of the plant based on satisfactory surveillance tests and no previous failures. Therefore, the exposed time for the de-energized hydrogen igniters was less than 3 days, resulting in very low safety significance. The Appendix H evaluation and the final risk significance determination were reviewed and concurred on by a regional senior reactor analyst. This issue is a latent issue associated with original plant equipment and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 1R15). Inspection Report# : 2011004 (pdf) Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform Preventative Maintenance on the Fuel Handling Bridge Paddle Switch Green. The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for inadequate preventative maintenance instructions, which resulted in the loss of control of the fuel handling bridge in the spent fuel pool. On July 15, 2011, while moving spent fuel from the spent fuel pool to the dry cask loading pool, the fuel handling platform did not stop when the paddle switch was released from the reverse position. The paddle switch did not return to the neutral position as designed, and the bridge continued to move in the reverse direction. The fuel handling bridge tripped the zone limit switches and came to a

stop. The licensee concluded that the switches had to be cleaned, adjusted, and re-greased periodically to ensure proper operation. Immediate corrective actions included replacing the paddle switch and revising the preventive maintenance instruction to clean and re-grease the paddle switch before every dry cask fuel campaign. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2011-04896. The finding is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases caused by accidents or events. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not result from fuel handling errors that caused damage to fuel clad integrity because the fuel handling bridge movement was arrested prior to coming in contact with the spent fuel pool wall. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with the operational experience component because the licensee failed to evaluate the need to update the preventative maintenance procedure for known issues associated with the fuel handling bridge paddle switch prior to the implementation of the dry fuel storage campaign [P.2(b)] (Section 4OA2). Inspection Report# : 2011004 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustible Stored in the Fire Exclusion Zone Near the Independent Spent Fuel Storage Installation Green . The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that transient combustible were not stored in the fire exclusion zone near the independent spent fuel storage installation. The inspectors performed a quarterly fire protection inspection of independent spent fuel storage installation and identified a large air conditioner with combustible material covering it located in the fire exclusion zone that was within 60 feet of the dry fuel storage pad. The inspectors determined through interviews that the material had been placed there the previous day by the maintenance department. As immediate corrective action the licensee removed the combustible material from the area. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00455. This finding was more than minor because it was associated human performance attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire prevention and administrative controls category. The inspectors assigned a low degradation rating due to the fact that the amount of combustible material in the area was minimal. The inspectors concluded that the finding was of very low safety significance (Green) due to the fact there were no fire ignition sources in the area. The cause of this finding has a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively communicate expectations regarding storage of combustible material near the dry fuel storage pad. [H.4(b)] (Section 1R05.1.b) Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Design Control for the Mitigation Monitoring System Modification Green. The inspectors reviewed a self-revealing, Green finding of EN-DC-115, Engineering Change Process, involving the failure to maintain adequate design control measures associated with the installation of the mitigation monitoring system. On November 8, 2010, a reactor coolant pressure boundary failure occurred at the skid mounted Online Noble Chemical - Mitigation Monitoring System pump inside primary containment. The positive displacement sample pump ejected the pump piston from the housing, resulting in an approximate 7 gpm leak of reactor coolant. The steam leak resulted in a reactor recirculation system flow control valve lockup (due to hydraulic power unit motor failure) and approximately 15,000 square feet of contaminated area in the primary containment structure. The licensee failed to ensure proper validation testing for the pump prior to installation.

Specifically, the licensee did not ensure that the pump could withstand the operating pressures and temperatures of the system in which it was installed. The licensee removed the mitigation monitoring system from service and isolated the skid from the reactor water cleanup system. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2010-07852. The finding is more than minor because it affects the design control attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Therefore, using inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet for LOCA initiators, the inspectors concluded that the finding was of very low safety significance (Green) because the failure of the mitigation monitoring system would not have exceeded technical specifications limits for identified leakage in the reactor coolant system. This finding has a crosscutting aspect in the work practices component of the human performance area; because the licensee failed to adequately oversee the design of the mitigation monitoring system such that nuclear safety is supported. [H.4(c)] (Section 4OA3.2.b) Inspection Report# : 2011002 (pdf) Emergency Preparedness Significance: Dec 31, 2010 Identified By: NRC Item Type: VIO Violation Failure to Have Guidelines for the Choice of Protective Actions During an Emergency Consistent with Federal Guidance Green. A cited violation of 10 CFR 50.47(b)(10) was identified for failure to develop and have in place guidelines for the choice of protective actions during an emergency that were consistent with federal guidance. Federal guidance for the choice of protective actions during an emergency is described in EPA-400-R-92-001 and states, in part, that evacuation is seldom justified when doses are less than protective action guides. The licensees automatic process that extended existing protective action recommendations with changes in wind direction without considering radiation dose was identified as a performance deficiency. This finding is more than minor because it affects the Emergency Preparedness Cornerstone objective of implementing adequate measures to protect the health and safety of the public during a radiological emergency, and is associated with the cornerstone attributes of emergency response organization performance and procedure quality. This finding was determined to be of very low safety significance because it was a failure to comply with NRC requirements, was associated with risk significant planning standard 10 CFR 50.47(b)(10), and was not a risk significant planning standard functional failure or a planning standard degraded function. The finding was not a functional failure or degraded planning standard function because appropriate protective action recommendations for the public would have been made for all areas where protective action guides were exceeded. This finding is a cited violation of 10 CFR 50.47(b)(10) because the licensee failed to restore compliance with NRC requirements in a timely manner. The finding is related to the corrective action element of the problem identification and resolution crosscutting aspect because the licensee failed to take corrective actions to address the safety issue in a timely manner [P1.d] (Section 1EP5) Inspection Report# : 2010005 (pdf) Occupational Radiation Safety Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a

locked high radiation area Green. The inspectors identified a noncited violation of Technical Specification 5.7.2, resulting from the licensees failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area. The finding was placed into the corrective action program as Condition Report CR-GGN-2011-01045, and corrective action was being evaluated. The failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area is a performance deficiency. The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to use qualified radiation protection technicians to provide job coverage in a high radiation area with dose rates in excess of 1000 mrem/hr had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. (Section 2RS01.b) Inspection Report# : 2011002 (pdf) Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : January 04, 2012

Grand Gulf 1 4Q/2011 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions for Reactor Core Isolation Cooling System Venting Green. The team identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct an inadequate venting procedure for the reactor core isolation cooling system. Corrective actions were not taken in a timely enough manner such that resolution was reached prior to time to demonstrate the licensee met their applicable technical specification surveillance requirement. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-07669 and subsequently altered their procedure, which performs the technical specification surveillance requirement to demonstrate that it meets the applicable requirements. This finding is more than minor because it affects the procedure quality 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. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the decision making component. The licensee did not use conservative assumptions when deciding to pursue corrective action for venting of the reactor core isolation cooling system piping to demonstrate their action was safe in order to proceed rather than demonstrating it was unsafe to disapprove the action [H.1(b)]. (Section 4OA2.5a) Inspection Report# : 2011006 (pdf) Significance: SL-IV Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit a Licensee Event Report for an Inoperable Reactor Core Isolation Cooling System Severity Level IV. The team identified a Severity Level IV noncited violation of 10 CFR 50.73, Licensee Event Report System, associated with the licensees failure to submit a licensee event report within 60 days following discovery of an event meeting the reporting criteria as specified. Specifically, the licensee was not meeting the technical specification surveillance requirement for venting the reactor core cooling isolation system and subsequently the system was inoperable in excess of the allowed outage time which constituted a condition prohibited by technical specifications. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8890. This finding affects the mitigating systems cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRCs ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section 6.9 of the Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect, as it is not indicative of current performance (Section 4OA2.5b).

Inspection Report# : 2011006 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Document a Condition as a Significant Condition Adverse to Quality Green. The team identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to identify and document a significant condition adverse to quality and report the condition to appropriate levels of management. As a result, a root cause analysis was not performed and more comprehensive actions to prevent recurrence were not considered for the condition. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011- 07671, to address the problem. This finding is more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance (Green) because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensees procedures for significant conditions adverse to quality were not complete and accurate enough to prevent the condition. [H.2(c)]. (Section 4OA2.5c) Inspection Report# : 2011006 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: VIO Violation Inadequate Corrective Action for a Leak on the Division II Emergency Diesel Generator Lube Oil Sump Green. The team identified a Green cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a leak on the Division II emergency diesel generator lube oil sump. Despite identification of the leak in 2004, ineffective attempts to repair the leak and previous identification by the NRC in 2009, the licensee dispositioned the leak as accept as-is without a full understanding of the lube oil sump leak and potential consequences. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8880. The condition was discovered and documented by the licensee in 2004. This finding was initially determined by the NRC to be a minor violation in 2009. Paragraph F of Section 2.10 of the NRC Enforcement Manual states in part that where a licensee does not take corrective action for a minor violation, the matter should be considered more than minor and associated with a green inspection finding and dispositioned in a cited or noncited violation, as appropriate. This finding is now determined to be more than minor because if left uncorrected the failure to restore the lube oil sump for the Division II emergency diesel generator to design conditions would have the potential to lead to a more significant safety concern, specifically, the leak could worsen and potentially affect operability of the emergency diesel generator. Due to the licensees failure to restore compliance within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy. This finding affects the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate this problem such that the resolutions addressed the causes [P.1(c)]. (Section 4OA2.5d) Inspection Report# : 2011006 (pdf) Significance: Sep 27, 2011

Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance for Components Used in Critical Applications Green. The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1.a for the licensees failure to follow a procedure that required them to evaluate components of critical systems in order to establish a preventive maintenance strategy, which resulted in unscheduled unavailability of safety-related systems and associated unscheduled entries into 72-hour shutdown Technical Specification action statements. The inspectors noted the following two examples dealing with failures of safety related equipment, which resulted in entering into shutdown limiting condition of operation. On June 2, 2011, Grand Gulf Nuclear Station experienced a failure of a relay in the standby service water B pump house ventilation system, which rendered the standby service water B system inoperable. The immediate corrective actions were to replace the relays and to restore the ventilation system. On June 22, 2011, the station experienced a failure of a motor contactor coil on breaker 52-154128, which caused the engineered safety feature electrical switchgear room cooler fan coil unit 1T46B003A not to run. The maintenance personnel determined the failure was due to a burnt motor contactor coil. The immediate corrective action was to replace the contactor coil and restore the room cooler. In both cases, the failed equipment was original plant equipment and preventive maintenance measures had not been established. The licensee entered these issues into the corrective action program as Condition Reports CR-GGN-2011-3730 and CR-GGN-2011-4313. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function of a single train for more than its technical specifications allowed outage time. This issue is a latent issue associated with original plant equipment and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 4OA2). Inspection Report# : 2011004 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Provide An Adequate Alternative Shutdown Procedure

  • Green. The team identified a noncited violation of License Condition 2.C(41), Fire Protection Program, for failing to ensure that the postfire safe shutdown procedure for fires requiring control room evacuation could be performed within the critical times required by the approved fire protection program. Specifically, two crews of operators simulating performance of Procedure 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Revision 036, did not give priority to the required safe shutdown components which are protected against fire damage and did not complete the equipment alignments within the times required by the thermal-hydraulic analysis. The team confirmed at the end of each walkdown that the operators involved did not know what the credited shutdown equipment was for a postfire safe shutdown or the critical time limits to be met. The team also confirmed that the licensee had not performed timed walkdowns to validate that the procedure would complete the required actions for postfire safe shutdown within the times required by the thermal-hydraulic analysis. The licensee entered this into their corrective action program as CR GGN 2011 02721, implemented compensatory measures to focus the operators priority on the required safe shutdown components and implemented a procedure revision.

The failure to provide an adequate procedure to implement the requirements of the approved fire protection program for a fire in the control room is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. The scenario impacted operators being ready to emergency depressurize the reactor and reflood using a residual heat removal pump. Because a bounding change to core damage frequency was 4.13 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding is of very low safety significance (Green). The finding did not have a crosscutting aspect since the primary cause did not fit any crosscutting aspects. (Section 1R5.5.b.1)

Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Assure Equipment Required For Postfire Safe Shutdown Was Protected Against Fire Damage Green. The team identified a noncited violation of License Condition 2.C(41), Fire Protection Program, for failing to assure that equipment relied upon for safe shutdown following a fire in the control room was protected against fire damage. Specifically, Procedure 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Revision 036, relied on the automatic operation of and indications from the load shedding and sequencing system. The team identified that this system was not isolated from potential damage due to a fire in the control room and the procedure did not adequately address the potential that fire damage to the system could effect the postfire safe shutdown capability by spuriously starting or stopping electric loads. The licensee entered this into their corrective action program as CR GGN 2011 02721. The failure to assure that equipment required to successfully implement the safe shutdown procedure for a fire in the control room was protected against fire damage is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 1.97 x 10-8, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding did not have a crosscutting aspect since it was not indicative of current performance, in that the licensee had established the current procedure more than three years prior to this finding. (Section 1R5.5.b.2) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions to Protect Safe Shutdown Equipment From Fire Damage Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to take timely corrective action to modify the control circuits for 33 motor operated valves that are relied upon during safe shutdown due to fire. Noncited violation NCV 05000416/2008006-04, Failure to Ensure That Damage to Motor-Operated Valve Circuits Would Not Prevent Safe Shutdown, documented the licensees inadequate review of Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During Control Room Fire. The licensee failed to develop modification packages such that motor operated valve control circuit modifications could be implemented during the fall 2010 refueling outage. As a result, 33 motor operated valves associated with safe shutdown equipment continue to remain susceptible to potential damage during spurious operation due to circuit hot shorts. The licensee has maintained a fire watch as a compensatory measure. The licensee entered this into their corrective action program as CR GGN 2011 02779. The failure to take timely corrective actions to address the potential for fire induced hot shorts to impact the ability to safely shutdown the plant following a fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 9.58 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding had a crosscutting aspect in the area of Human Performance associated with Decision Making, because the licensee failed to demonstrate that nuclear safety is an overriding priority. Specifically, the licensee did not promptly initiate control

circuit reviews and implement modifications required for corrective actions after the licensees inadequate evaluation of Information Notice 92-18 was identified in the 2008 violation. [H.1(a)] (Section 1R5.6) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions To Assure Postfire Safe Shutdown Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for inadequate corrective actions to address the potential for fire induced hot shorts to impact the ability to trip a control rod group as described in Information Notice 2007-07. The licensees evaluation of Information Notice 2007-07 stated in part, provisions have been included in 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, to trip the proper reactor protective system breakers to ensure that the reactor scram occurs. The team noted that Procedure 05-1-02-II-1 contained a conditional statement for the operator to determine if opening the reactor protective system breakers is required. The procedure did not provide assurance that all control rod groups insert since the control room indications to be utilized by the operator were not identified and confirmed to be reliable during fires requiring control room evacuation. The licensee entered this finding into its corrective action program under CR-GGN-2011-02780, implemented compensatory measures to ensure the operators de-energized the reactor protection system, and implemented a procedure change. The failure to take adequate corrective actions to address the potential for fire induced hot shorts to impact the ability to safely shutdown the plant following a fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 9.58 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding did not have a crosscutting aspect since it was not indicative of current performance. The licensee had incorrectly assessed the applicability of Information Notice 2007-07 more than three years prior to this finding. (Section 4OA2.b) Inspection Report# : 2011007 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation Door Seals Protecting Safety Related Equipment Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. Inspectors found the entrance door to the diesel generator building and the entrance door to the division 2 diesel generator in a degraded condition. The inspectors identified that the door seals did not make complete contact with the door frames all the way around as required by procedure. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-02575. The finding is more than minor because it is associated with the protection against external factors attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multiple trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis. The result was a delta-core damage frequency of 3.3E-7/yr and a delta-large early release frequency of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The inspectors determined the apparent cause of this finding was

that licensee personnel were not adequately trained to perform these inspections. Therefore this finding has a cross-cutting aspect in the area of human performance associated with resources in that the licensees training of personnel was not adequate in performing inspection of the probable maximum precipitation door seals [H.2(b)](Section 1R01). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Scaffold Control Procedure Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement scaffolding control procedural requirements related to post-installation inspections and engineering safety evaluations for scaffolding constructed within 2 inches of safety-related or fire protection equipment. During plant walkdowns, inspectors identified multiple examples of the licensee not properly implementing Entergys corporate and site procedures for the control of scaffolding. The licensees immediate corrective actions included inspecting the scaffolding that had been installed, modifying or removing it where appropriate, and properly posting the scaffolds. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2011-03480, CR-GGN-2011-03601, CR-GGN-2011-03602, and CR-GGN-2011-03603. The inspectors determined that this finding is more than minor because it is associated with the external factors and equipment performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green), because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the apparent cause of this finding was lack of supervisor oversight during scaffold construction. Therefore the finding has a cross-cutting aspect in the area of human performance associated with work practices, in that the licensee did not provide effective supervisor oversight of workers constructing scaffolding to ensure these activities were performed per procedural requirements [H.4(c)](Section 1R04). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Conditions Adverse to Fire Protection Green. The inspectors identified a noncited violation of License Condition 2.C(41) for the failure to identify conditions adverse to the fire protection program. Specifically, during required inspections of the material condition of the sprinkler system, the licensee failed to identify several instances of bent or misaligned sprinkler head deflector plates and a painted sprinkler head. Corrective action included correcting bent or misaligned plates and replacing the painted sprinkler head. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-03132. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the safety concern is that the number of bent or misaligned sprinkler heat canopies and painted sprinkler heads would not provide an adequate area-wide coverage of suppression. The inspectors evaluated the significance of this finding using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The deficiency involved the Fixed Fire Protection Systems category. Using Appendix F, Enclosure , "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," the inspectors determined that the deficiency had low degradation since less than 10 percent of the heads in the affected fire area were nonfunctional, a functional head remained within 10 feet of the combustibles of concern, and the system remained nominally code compliant. This finding screened as having very low safety significance (Green) in Phase 1of Manual Chapter 0609, Appendix F. This finding has a cross-cutting aspect in the area of human performance associated with resources because the procedure used to inspect the condition of these sprinklers did not contain

specific criteria for identifying unacceptable sprinkler conditions [H.2(c)](Section 1R05). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure that Safety Related Manholes were Properly Sealed to Prevent the Entry of Flammable Liquid Green. The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that manholes MH01, MH20 and MH21 were properly sealed to prevent the entry of flammable liquid. During the performance of the manhole/vault inspection, the inspectors were reviewing engineering change packages associated with solar sump pumps for MH20 and MH21. During their review, they determined that the licensee was not meeting the requirements of their license bases documents for MH20 and MH21, which contain safe shutdown cables for standby service water trains A and B. The licensees immediate corrective action included placing hazmat barricades around each manhole to prevent flammable fluids from entering the manholes. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00562. This finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 3b, Item 1 directs the inspectors to Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. However, an NRC senior reactor analyst determined that the unique nature of this performance deficiency did not lend itself to analysis by the methods provided in Appendix F. Therefore, a Phase 3 analysis was performed. Based on a bounding analysis, the analyst determined that the change in core damage frequency was approximately 1.5E-7/yr. The result was low because of the relatively short periods of time that fuel was actually being transferred, the low probability of transfer system failures, and the low likelihood that a loss of normal service water initiator would occur following a fire in the subject manholes. This noncited violation was therefore determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to initiate a condition report when the issue was identified during the development of their engineering change package, which resulted in the failure to ensure the safety related manholes were sealed in accordance with their license based documents [P.1(a)](Section 1R06). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Provide Adequate Procedures for High Pressure Core Spray Minimum Flow Valve Surveillance Testing Green. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for the licensees failure to provide adequate testing procedures, which resulted in the high pressure core spray minimum flow valve inadvertently stroking approximately 11 times during a surveillance test. The excessive stroking of the valve resulted in the unplanned inoperability of the high pressure core spray system because the valves feeder breaker overcurrent instantaneous trip setpoint had drifted below the manufacturers tolerance for the existing setting. As immediate corrective action, the licensee replaced the degraded breaker. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01901. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function since the high pressure core spray system would still have been functional even with the minimum flow valve potentially failing open. Additionally, it did not represent a loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with operating experience in that licensee had not incorporated operating experience from a similar event that had occurred at another

Entergy site [P.2(b)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Loose Fuse Clips in Division 3 Emergency Diesel Generator Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take adequate corrective actions for a significant condition adverse to quality associated with the division 3 emergency diesel generator. While performing a maintenance effectiveness review of the diesel generators, the inspectors noted on October 17, 2009, at 9:07 p.m., the FU-7 fuse for the division 3 diesel generator was determined to have a faulty fuse clip, resulting in the inoperability of the diesel generator due to loss of power to the direct current powered fuel pumps. Then on March 18, 2011, the division 3 emergency diesel generator was again rendered inoperable due to a faulty fuse clip on the FU-8 fuse holder, which is of the same design and function as the FU-7 fuse holder in the previous occurrence. Short term corrective action included replacing the fuse holder. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01868. The finding is more than minor because it is associated with equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, 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 in the area of human performance associated with resources because the training provided to correct the initial event was not adequate to ensure proper fuse installation and verify good connection existed between the fuse and fuse holder [H.2(b)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Assure Configuration Control of Safety Related Systems Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to review the suitability of leaving test fittings on reactor coolant system flow transmitter equalizing block drain ports instead of the design specified manifold plugs. As corrective action, the licensee replaced the test fittings with the correct drain plugs. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-04485. This finding is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability of functionality, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, associated with work practices, because the licensee failed to ensure that human error prevention techniques, such as holding pre-job briefings, self- and peer-checking, and proper documentation of activities were utilized such that work activities were performed safely and personnel did not proceed in the face of uncertainty or unexpected circumstances Specifically, the licensee failed to review the suitability of installing test and brass fittings on pressure, differential pressure and flow transmitter block valve drain ports instead of the design specified manifold plugs. [H.4(a)](Section 1R12). Inspection Report# : 2011003 (pdf)

Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow a Procedure Resulting in the Inoperability of the Reactor Core Isolation Cooling System Primary Containment Isolation Valve Green. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, for failure to follow a procedure resulting in the inoperability of the reactor core isolation cooling system primary containment isolation valve. This occurred while the licensee was performing surveillance on the reactor core isolation cooling system and incorrectly attached a jumper to the wrong terminal point resulting in blowing a fuse that caused a loss of control power to the reactor core isolation cooling primary containment isolation valve 1E51-F031. As immediate corrective action, the licensee removed the jumper and replaced the control power fuse. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01932. The finding is more than minor since it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. In addition, this finding had a human performance cross-cutting aspect associated with work practices in that the licensee did not use the proper human performance techniques of self-checking to prevent the loss of control power to a primary containment isolation valve [H.4(a)](Section 1R22). Inspection Report# : 2011003 (pdf) Significance: SL-IV Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update Available Low Pressure Coolant Injection Loops in the Updated Final Safety Analysis Report SLIV. Inspectors identified a noncited violation of 10 CFR 50.71(e)(4), which requires the final safety analysis report be updated, at intervals not exceeding 24 months, to reflect changes made in the facility or procedures described in the final safety analysis report. Licensee personnel failed to update the original revision of the final safety analysis report to reflect the actual number of low pressure coolant injection loops available for automatic initiation during shutdown cooling operations in Mode 3. The licensee plans to update the final safety analysis report at the next scheduled revision. This finding was entered into the licensees corrective action program as condition report CR-GGN-2011-01631. The failure of licensing personnel to update the final safety analysis report to reflect the available low pressure coolant injection loops for automatic initiation during shutdown cooling operations in Mode 3 was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy, dated September 30, 2010, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Demonstrate Maintenance Effectiveness of Train B Control Room Air Conditioner Green. The inspectors identified a noncited violation of 10 CFR Part 50.65(a)(2) for the licensees failure to demonstrate that the performance of the train B control room air conditioner was being effectively controlled through the performance of appropriate preventive maintenance. Engineering did not properly evaluate maintenance rule functional failures resulting in the system remaining in an a(2) status instead of an a(1) status. As corrective action, the train B control room air conditioner was moved into an a(1) status. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2011-01623. The finding was more than minor because it was associated with the equipment performance attribute of the

Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the maintenance rule aspect of the finding did not cause an actual loss of safety function of the system nor did it cause a component to be inoperable. As corrective action, the train B control room air conditioner was moved into an (a)(1) status. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because licensee personnel failed to make appropriate safety-significant or risk-significant decisions to address the multiple failures of the train B control room air conditioner compressor. [H.1(a)] (Section 1R12.b.2) Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Prevent Recurrence of Control Room Air Conditioner Compressor Tripping Due to Low Oil Pressure Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, after the licensee failed to determine the cause and prevent recurrence of a significant condition adverse to quality associated with the train B control room air conditioner compressor tripping due to low oil pressure. Specifically, on December 13, 2010, the train B control room air conditioner compressor tripped on low oil pressure after the licensee had performed a root cause analysis to identify the cause and prevent recurrence of a similar compressor trip on October 14, 2010. As immediate corrective action, the licensee installed an inline suction filter. No additional failures have occurred since its installation. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2010-07315. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheets, the inspectors determined that a Phase 2 analysis was required because the finding represented a loss of system safety function. The plant-specific risk informed notebook does not include the evaluation of risk caused by the loss of cooling to the main control room. Therefore, the senior reactor analyst conducted a Phase 3 analysis. Based on the bounding analysis, the analyst determined that the change in core damage frequency result was 5.9 x 10-7. This noncited violation was therefore determined to be of very low safety significance (Green). This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because licensee personnel failed to thoroughly evaluate the multiple failures of the train B control room air conditioner compressor. [P.1(c)] (Section 4OA3.1.b) Inspection Report# : 2011002 (pdf) Barrier Integrity Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Ensure Correct Fuses were Installed in the Hydrogen Igniter Control Circuits Green. The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to install the correct amperage fuses in the hydrogen igniter control circuit. On August 4, 2011, the inspectors were performing an operability review of a condition report dealing with the division 1 hydrogen igniters. The licensee had determined that half the division 1 hydrogen igniters would not energize, and in their investigation, they determined that the loss of power to the hydrogen igniters was due to a blown fuse. The licensee also determined that the blown fuse was 0.3 amps and should have been 0.8 amps per plant drawings. The licensee performed an operability determination for the as found condition and determined that the circuit required 0.193 amps to power the circuit, which included the light bulbs. The inspectors reviewed the operability determination and the calculations and determined that the licensees conclusions were reasonable. The licensee immediate

corrective action was to replace the incorrect fuses one division at a time with the correct size 0.8 amp fuses and restore the hydrogen igniters to operable status. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-005388. This finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone, and it adversely affected the cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that Appendix H, Containment Integrity Significance Determination Process, was required. Inspectors determined that this was a type B finding and, using section 6.0, determined that the finding was of very low safety significance (Green) because during their review, the inspectors noted that the hydrogen igniters had maintain functionality over the life of the plant based on satisfactory surveillance tests and no previous failures. Therefore, the exposed time for the de-energized hydrogen igniters was less than 3 days, resulting in very low safety significance. The Appendix H evaluation and the final risk significance determination were reviewed and concurred on by a regional senior reactor analyst. This issue is a latent issue associated with original plant equipment and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 1R15). Inspection Report# : 2011004 (pdf) Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform Preventative Maintenance on the Fuel Handling Bridge Paddle Switch Green. The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for inadequate preventative maintenance instructions, which resulted in the loss of control of the fuel handling bridge in the spent fuel pool. On July 15, 2011, while moving spent fuel from the spent fuel pool to the dry cask loading pool, the fuel handling platform did not stop when the paddle switch was released from the reverse position. The paddle switch did not return to the neutral position as designed, and the bridge continued to move in the reverse direction. The fuel handling bridge tripped the zone limit switches and came to a stop. The licensee concluded that the switches had to be cleaned, adjusted, and re-greased periodically to ensure proper operation. Immediate corrective actions included replacing the paddle switch and revising the preventive maintenance instruction to clean and re-grease the paddle switch before every dry cask fuel campaign. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2011-04896. The finding is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases caused by accidents or events. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not result from fuel handling errors that caused damage to fuel clad integrity because the fuel handling bridge movement was arrested prior to coming in contact with the spent fuel pool wall. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with the operational experience component because the licensee failed to evaluate the need to update the preventative maintenance procedure for known issues associated with the fuel handling bridge paddle switch prior to the implementation of the dry fuel storage campaign [P.2(b)] (Section 4OA2). Inspection Report# : 2011004 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustible Stored in the Fire Exclusion Zone Near the Independent Spent Fuel Storage Installation Green . The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that transient combustible were not stored in the fire exclusion zone near the independent spent fuel storage installation. The inspectors performed a quarterly fire protection inspection of independent spent fuel storage installation and identified a large air conditioner with combustible material covering it located in the fire exclusion zone that was within 60 feet of the dry fuel storage pad. The inspectors determined through interviews that the

material had been placed there the previous day by the maintenance department. As immediate corrective action the licensee removed the combustible material from the area. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00455. This finding was more than minor because it was associated human performance attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that the finding impacted the fire prevention and administrative controls category. The inspectors assigned a low degradation rating due to the fact that the amount of combustible material in the area was minimal. The inspectors concluded that the finding was of very low safety significance (Green) due to the fact there were no fire ignition sources in the area. The cause of this finding has a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to effectively communicate expectations regarding storage of combustible material near the dry fuel storage pad. [H.4(b)] (Section 1R05.1.b) Inspection Report# : 2011002 (pdf) Significance: Mar 27, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Design Control for the Mitigation Monitoring System Modification Green. The inspectors reviewed a self-revealing, Green finding of EN-DC-115, Engineering Change Process, involving the failure to maintain adequate design control measures associated with the installation of the mitigation monitoring system. On November 8, 2010, a reactor coolant pressure boundary failure occurred at the skid mounted Online Noble Chemical - Mitigation Monitoring System pump inside primary containment. The positive displacement sample pump ejected the pump piston from the housing, resulting in an approximate 7 gpm leak of reactor coolant. The steam leak resulted in a reactor recirculation system flow control valve lockup (due to hydraulic power unit motor failure) and approximately 15,000 square feet of contaminated area in the primary containment structure. The licensee failed to ensure proper validation testing for the pump prior to installation. Specifically, the licensee did not ensure that the pump could withstand the operating pressures and temperatures of the system in which it was installed. The licensee removed the mitigation monitoring system from service and isolated the skid from the reactor water cleanup system. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2010-07852. The finding is more than minor because it affects the design control attribute of the Barrier Integrity Cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Therefore, using inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet for LOCA initiators, the inspectors concluded that the finding was of very low safety significance (Green) because the failure of the mitigation monitoring system would not have exceeded technical specifications limits for identified leakage in the reactor coolant system. This finding has a crosscutting aspect in the work practices component of the human performance area; because the licensee failed to adequately oversee the design of the mitigation monitoring system such that nuclear safety is supported. [H.4(c)] (Section 4OA3.2.b) Inspection Report# : 2011002 (pdf) Emergency Preparedness Occupational Radiation Safety Significance: Mar 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area

Green. The inspectors identified a noncited violation of Technical Specification 5.7.2, resulting from the licensees failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area. The finding was placed into the corrective action program as Condition Report CR-GGN-2011-01045, and corrective action was being evaluated. The failure to use a qualified radiation protection technician to provide direct continuous coverage of work in a locked high radiation area is a performance deficiency. The finding was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, the failure to use qualified radiation protection technicians to provide job coverage in a high radiation area with dose rates in excess of 1000 mrem/hr had the potential to increase personnel dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to have very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. (Section 2RS01.b) Inspection Report# : 2011002 (pdf) Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : March 02, 2012

Grand Gulf 1 1Q/2012 Plant Inspection Findings Initiating Events Significance: Mar 23, 2012 Identified By: Self-Revealing Item Type: FIN Finding Manual Reactor Scram Caused by Failure to Ensure the Main Steam Supply Valve to Reactor Feed Pump Turbine B was Full Open Green . The inspectors reviewed a Green self-revealing finding for the failure to ensure the correct position (full open) of the main steam supply valve 1N11-F014B to reactor feed pump turbine B, which resulted in a manual reactor scram due to decreasing reactor water level. During plant shutdown activities to begin refueling outage 18, the at-the-controls operator manually scrammed the reactor from approximately 23 percent rated thermal power due to the decreasing reactor water level. Water level in the reactor was decreasing because valve 1N11-F014B was not fully open, and because pressure in the main steam lines had been reduced when the crew opened turbine bypass valves to begin cooling the main turbine. With valve 1N11-F014B less than fully open and reduced steam pressure, the operating feed pump wasnt able to maintain water level. After the scram, reactor core isolation cooling and reactor feed pump turbine A were used to restore water level. The licensee plans to repair valve 1N11-F014B during the current refuelling outage. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01838. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors concluded that the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment would not be available. The inspectors, in consultation with the regional senior reactor analyst, performed a Phase 2 estimation using the pre-solved work sheets for Grand Gulf Nuclear Station. The inspectors determined by entering the power conversion system column that the finding was of very low safety significance (Green). This result was validated by the senior reactor analyst using the current revision of the plant-specific SPAR model. The inspectors determined the finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the operating staff proceeded with the start up of the reactor feed pump B with the main steam supply valve 1N11-F014B in an unknown position [H.1(b)](Section 1R11). Inspection Report# : 2012002 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Online Risk Assessment Per Severe Weather Off Normal Procedure Due to a Declared Tornado Warning Affecting Grand Gulf Nuclear Station Green. The inspectors identified a Green non-cited violation of Technical Specifications 5.4.1.a for the failure to perform an online risk assessment per severe weather off normal procedure due to a declared tornado warning affecting Grand Gulf Nuclear Station. At 7:41 p.m., on February 15, 2012, the National Weather Service issued a tornado warning for Claiborne County, the county in which Grand Gulf Nuclear Station is located. In response to a tornado warning, licensee procedures required them to enter Off-Normal Operating Procedure 05-1-02-VI-2, Severe Weather, and evaluate online risk. This severe weather condition would have resulted in the licensee entering into an orange risk condition. On February 16, 2012, the inspectors identified that the licensee had not made a log entry for entry into their off normal severe weather procedure during the preceding evening and therefore had not evaluated online risk status for the severe weather condition. In response to the inspectors observations, the licensee initiated a condition report detailing the failures to enter the off normal procedure and enter the correct risk condition. The licensee has implemented short-term corrective actions to ensure the site adequately evaluates the risk associated with

adverse weather. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01707. The finding is more-than-minor because it is associated with the Initiating Events Cornerstone attribute of protection against external events, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Appendix K; Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1; Assessment of Risk Deficit; and consulting with the regional senior reactor analyst, the inspectors determined the finding to be of very low safety significance based on a licensees calculated determination of the incremental core damage probability deficit of 4.0E-08. This result was validated by the senior reactor analyst using the current revision of the plant-specific SPAR model. The inspectors determined the finding has a cross-cutting aspect in the area of human performance associated with the resources component because the on-shift senior reactor operators did not have adequate access to current weather information that would prompt control room personnel to re-evaluate risk due to changing weather conditions [H.2(d)](Section 1R13). Inspection Report# : 2012002 (pdf) Mitigating Systems Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions Associated with Division 1 and 2 Standby Service Water Safety Related Cables that were Partially Submerged in Cable Manhole/Vault Green . The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take timely corrective actions to correct a condition adverse to quality associated with division 1 and 2 standby service water safety related cables that were partially submerged in a cable manhole/vault. The inspectors reviewed work order 52284535 and noted that the sump pump for manhole/vault MH-01, which contained standby service water cables for division 1 and 2, was determined to be non-functional on September 10, 2011. The inspectors determined that a work order to repair the non-functioning sump pump had been developed but that the work order had not yet been scheduled. During a subsequent inspection, manhole/vault MH-01 was found to contain approximately three feet of water, with water partially covering some of the safety related cables. The electricians immediately pumped manhole/vault MH-01 and wrote a condition report. The licensee repaired the sump pump the next week and declared it functional. The cables remained operable based on the results of meggar tests. The licensee entered this issue into their corrective action program as condition reports CR-GGN-2012-00503, 01324, and 01389. The finding is more than minor because it is associated with the equipment performance attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent either a loss of system safety function or an actual loss of safety function of a single train of one or more non-Technical Specification 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 in the area of human performance associated with the work practices component, in that the licensee personnel did not initiate a condition report as required by licensee procedure when the work order associated with sump pump testing of MH-01 determined that the sump pump was not functioning properly [H.4(b)] (Section 1R06). Inspection Report# : 2012002 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation

Failure to Correct a Condition Adverse to Fire Protection, in That the Licensee Failed to Adequately Provide Contingency Lighting in the Fire Brigade Dress Out Area While Normal Lighting was Inoperabl Green. The inspectors identified a Green non-cited violation of Facility Operating License Condition 2.C(41), for the failure to correct a condition adverse to fire protection. Specifically, the licensee failed to adequately provide contingency lighting in the fire brigade dress out area while normal lighting was inoperable due to maintenance on an associated breaker. The inadequate lighting delayed fire brigade response to a potential fire in the turbine building. Immediate corrective action included placing temporary lighting in the area. Normal lighting to the area was restored the next week. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01488. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined from table 3b that issues related to performance of the fire brigade are not included in Appendix F and require NRC management review using Appendix M. Regional management review evaluated the overall impact of lighting issue in the fire brigade dress out area and concluded that, while the fire protection defense-in-depth was affected by the performance deficiency, the overall defense-in-depth of the front-line systems was not impacted because of train separation and safe shutdown analysis at the site. Therefore the finding screened as having very low safety significance (Green) in accordance with Manual Chapter 0609, Appendix M. The inspectors determined the finding had a cross-cutting aspect in the area of human performance associated with the work control component, in that licensee personnel failed to ensure adequate job site conditions (lighting in the fire bridge dress out area) were in place prior to performance electrical maintenance in the turbine building [H.3(a)] (Section 40A3). Inspection Report# : 2012002 (pdf) Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation Door Seals Protecting Safety Related Equipment Green. The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. Inspectors found that one of the door seals to standby service water pump house A was in a degraded condition. The inspectors identified that the door seal did not make complete contact with the door frame all the way around. The licensee determined that the probable maximum precipitation seal for the identified door was in a degraded condition. Failure of this door seal during a probable maximum precipitation event could potentially cause flooding of the standby service water pump house A. Immediate corrective actions included the site initiating compensatory actions for the degraded seal by staging sand bags in the area and requiring monitoring of the affected door during heavy rainfall. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2011-07687. The finding is more than minor because it is associated with the protection against external factors attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors used the seismic, flooding, and severe weather Table 4b and determined that it would not affect multiple trains of safety equipment and that the finding had very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the resources component in that the licensees procedure used for the inspection of the door seals did not take into account the status of the pump house ventilation system while performing the door seal inspection, and therefore, the licensee failed to make the required adjustments to the door seals resulting in their inspections of the probable maximum precipitation door seals being inadequate [H.2(c)] (Section 1R05). Inspection Report# : 2011005 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions for Reactor Core Isolation Cooling System Venting

Green. The team identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct an inadequate venting procedure for the reactor core isolation cooling system. Corrective actions were not taken in a timely enough manner such that resolution was reached prior to time to demonstrate the licensee met their applicable technical specification surveillance requirement. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-07669 and subsequently altered their procedure, which performs the technical specification surveillance requirement to demonstrate that it meets the applicable requirements. This finding is more than minor because it affects the procedure quality 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. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the decision making component. The licensee did not use conservative assumptions when deciding to pursue corrective action for venting of the reactor core isolation cooling system piping to demonstrate their action was safe in order to proceed rather than demonstrating it was unsafe to disapprove the action [H.1(b)]. (Section 4OA2.5a) Inspection Report# : 2011006 (pdf) Significance: SL-IV Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit a Licensee Event Report for an Inoperable Reactor Core Isolation Cooling System Severity Level IV. The team identified a Severity Level IV noncited violation of 10 CFR 50.73, Licensee Event Report System, associated with the licensees failure to submit a licensee event report within 60 days following discovery of an event meeting the reporting criteria as specified. Specifically, the licensee was not meeting the technical specification surveillance requirement for venting the reactor core cooling isolation system and subsequently the system was inoperable in excess of the allowed outage time which constituted a condition prohibited by technical specifications. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8890. This finding affects the mitigating systems cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRCs ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section 6.9 of the Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect, as it is not indicative of current performance (Section 4OA2.5b). Inspection Report# : 2011006 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Document a Condition as a Significant Condition Adverse to Quality Green. The team identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to identify and document a significant condition adverse to quality and report the condition to appropriate levels of management. As a result, a root cause analysis was not performed and more comprehensive actions to prevent recurrence were not considered for the condition. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011- 07671, to address the problem. This finding is more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance (Green) because it did not create a loss of system safety function of a single train for greater than the

technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensees procedures for significant conditions adverse to quality were not complete and accurate enough to prevent the condition. [H.2(c)]. (Section 4OA2.5c) Inspection Report# : 2011006 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: VIO Violation Inadequate Corrective Action for a Leak on the Division II Emergency Diesel Generator Lube Oil Sump Green. The team identified a Green cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a leak on the Division II emergency diesel generator lube oil sump. Despite identification of the leak in 2004, ineffective attempts to repair the leak and previous identification by the NRC in 2009, the licensee dispositioned the leak as accept as-is without a full understanding of the lube oil sump leak and potential consequences. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8880. The condition was discovered and documented by the licensee in 2004. This finding was initially determined by the NRC to be a minor violation in 2009. Paragraph F of Section 2.10 of the NRC Enforcement Manual states in part that where a licensee does not take corrective action for a minor violation, the matter should be considered more than minor and associated with a green inspection finding and dispositioned in a cited or noncited violation, as appropriate. This finding is now determined to be more than minor because if left uncorrected the failure to restore the lube oil sump for the Division II emergency diesel generator to design conditions would have the potential to lead to a more significant safety concern, specifically, the leak could worsen and potentially affect operability of the emergency diesel generator. Due to the licensees failure to restore compliance within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy. This finding affects the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate this problem such that the resolutions addressed the causes [P.1(c)]. (Section 4OA2.5d) Inspection Report# : 2011006 (pdf) Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance for Components Used in Critical Applications Green. The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1.a for the licensees failure to follow a procedure that required them to evaluate components of critical systems in order to establish a preventive maintenance strategy, which resulted in unscheduled unavailability of safety-related systems and associated unscheduled entries into 72-hour shutdown Technical Specification action statements. The inspectors noted the following two examples dealing with failures of safety related equipment, which resulted in entering into shutdown limiting condition of operation. On June 2, 2011, Grand Gulf Nuclear Station experienced a failure of a relay in the standby service water B pump house ventilation system, which rendered the standby service water B system inoperable. The immediate corrective actions were to replace the relays and to restore the ventilation system. On June 22, 2011, the station experienced a failure of a motor contactor coil on breaker 52-154128, which caused the engineered safety feature electrical switchgear room cooler fan coil unit 1T46B003A not to run. The maintenance personnel determined the failure was due to a burnt motor contactor coil. The immediate corrective action was to replace the contactor coil and restore the room cooler. In both cases, the failed equipment was original plant equipment and preventive maintenance measures had not been established. The licensee entered these issues into the corrective action program as Condition Reports CR-GGN-2011-3730 and CR-GGN-2011-4313. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating

Systems Cornerstone and adversely affected the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function of a single train for more than its technical specifications allowed outage time. This issue is a latent issue associated with original plant equipment and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 4OA2). Inspection Report# : 2011004 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Provide An Adequate Alternative Shutdown Procedure

  • Green. The team identified a noncited violation of License Condition 2.C(41), Fire Protection Program, for failing to ensure that the postfire safe shutdown procedure for fires requiring control room evacuation could be performed within the critical times required by the approved fire protection program. Specifically, two crews of operators simulating performance of Procedure 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Revision 036, did not give priority to the required safe shutdown components which are protected against fire damage and did not complete the equipment alignments within the times required by the thermal-hydraulic analysis. The team confirmed at the end of each walkdown that the operators involved did not know what the credited shutdown equipment was for a postfire safe shutdown or the critical time limits to be met. The team also confirmed that the licensee had not performed timed walkdowns to validate that the procedure would complete the required actions for postfire safe shutdown within the times required by the thermal-hydraulic analysis. The licensee entered this into their corrective action program as CR GGN 2011 02721, implemented compensatory measures to focus the operators priority on the required safe shutdown components and implemented a procedure revision.

The failure to provide an adequate procedure to implement the requirements of the approved fire protection program for a fire in the control room is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. The scenario impacted operators being ready to emergency depressurize the reactor and reflood using a residual heat removal pump. Because a bounding change to core damage frequency was 4.13 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding is of very low safety significance (Green). The finding did not have a crosscutting aspect since the primary cause did not fit any crosscutting aspects. (Section 1R5.5.b.1) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Assure Equipment Required For Postfire Safe Shutdown Was Protected Against Fire Damage Green. The team identified a noncited violation of License Condition 2.C(41), Fire Protection Program, for failing to assure that equipment relied upon for safe shutdown following a fire in the control room was protected against fire damage. Specifically, Procedure 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Revision 036, relied on the automatic operation of and indications from the load shedding and sequencing system. The team identified that this system was not isolated from potential damage due to a fire in the control room and the procedure did not adequately address the potential that fire damage to the system could effect the postfire safe shutdown capability by spuriously starting or stopping electric loads. The licensee entered this into their corrective action program as CR GGN 2011 02721. The failure to assure that equipment required to successfully implement the safe shutdown procedure for a fire in the

control room was protected against fire damage is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 1.97 x 10-8, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding did not have a crosscutting aspect since it was not indicative of current performance, in that the licensee had established the current procedure more than three years prior to this finding. (Section 1R5.5.b.2) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions to Protect Safe Shutdown Equipment From Fire Damage Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to take timely corrective action to modify the control circuits for 33 motor operated valves that are relied upon during safe shutdown due to fire. Noncited violation NCV 05000416/2008006-04, Failure to Ensure That Damage to Motor-Operated Valve Circuits Would Not Prevent Safe Shutdown, documented the licensees inadequate review of Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During Control Room Fire. The licensee failed to develop modification packages such that motor operated valve control circuit modifications could be implemented during the fall 2010 refueling outage. As a result, 33 motor operated valves associated with safe shutdown equipment continue to remain susceptible to potential damage during spurious operation due to circuit hot shorts. The licensee has maintained a fire watch as a compensatory measure. The licensee entered this into their corrective action program as CR GGN 2011 02779. The failure to take timely corrective actions to address the potential for fire induced hot shorts to impact the ability to safely shutdown the plant following a fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 9.58 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding had a crosscutting aspect in the area of Human Performance associated with Decision Making, because the licensee failed to demonstrate that nuclear safety is an overriding priority. Specifically, the licensee did not promptly initiate control circuit reviews and implement modifications required for corrective actions after the licensees inadequate evaluation of Information Notice 92-18 was identified in the 2008 violation. [H.1(a)] (Section 1R5.6) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions To Assure Postfire Safe Shutdown Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for inadequate corrective actions to address the potential for fire induced hot shorts to impact the ability to trip a control rod group as described in Information Notice 2007-07. The licensees evaluation of Information Notice 2007-07 stated in part, provisions have been included in 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, to trip the proper reactor protective system breakers to ensure that the reactor scram occurs. The team noted that Procedure 05-1-02-II-1 contained a conditional statement for the operator to determine if opening the reactor protective system breakers is required. The procedure did not provide assurance that all control rod groups insert since the control room indications to be utilized by the operator were not identified and confirmed to

be reliable during fires requiring control room evacuation. The licensee entered this finding into its corrective action program under CR-GGN-2011-02780, implemented compensatory measures to ensure the operators de-energized the reactor protection system, and implemented a procedure change. The failure to take adequate corrective actions to address the potential for fire induced hot shorts to impact the ability to safely shutdown the plant following a fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 9.58 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding did not have a crosscutting aspect since it was not indicative of current performance. The licensee had incorrectly assessed the applicability of Information Notice 2007-07 more than three years prior to this finding. (Section 4OA2.b) Inspection Report# : 2011007 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation Door Seals Protecting Safety Related Equipment Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. Inspectors found the entrance door to the diesel generator building and the entrance door to the division 2 diesel generator in a degraded condition. The inspectors identified that the door seals did not make complete contact with the door frames all the way around as required by procedure. The licensee initiated compensatory actions for the degraded seals, staging sand bags in the area and requiring monitoring of the affected doors during heavy rainfall. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-02575. The finding is more than minor because it is associated with the protection against external factors attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors used the seismic, flooding, and severe weather Table 4b and determined it would affect multiple trains of safety equipment. The inspectors consulted the regional senior reactor analyst, who performed a Phase 3 analysis. The result was a delta-core damage frequency of 3.3E-7/yr and a delta-large early release frequency of 6.6E-8/yr. These results confirmed that the finding had very low safety significance (Green). The inspectors determined the apparent cause of this finding was that licensee personnel were not adequately trained to perform these inspections. Therefore this finding has a cross-cutting aspect in the area of human performance associated with resources in that the licensees training of personnel was not adequate in performing inspection of the probable maximum precipitation door seals [H.2(b)](Section 1R01). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Scaffold Control Procedure Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement scaffolding control procedural requirements related to post-installation inspections and engineering safety evaluations for scaffolding constructed within 2 inches of safety-related or fire protection equipment. During plant walkdowns, inspectors identified multiple examples of the licensee not properly implementing Entergys corporate and site procedures for the control of scaffolding. The licensees immediate corrective actions included inspecting the scaffolding that had been installed, modifying or removing it where appropriate, and properly posting the scaffolds. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2011-03480, CR-GGN-2011-03601, CR-GGN-2011-03602, and CR-

GGN-2011-03603. The inspectors determined that this finding is more than minor because it is associated with the external factors and equipment performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green), because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the apparent cause of this finding was lack of supervisor oversight during scaffold construction. Therefore the finding has a cross-cutting aspect in the area of human performance associated with work practices, in that the licensee did not provide effective supervisor oversight of workers constructing scaffolding to ensure these activities were performed per procedural requirements [H.4(c)](Section 1R04). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Conditions Adverse to Fire Protection Green. The inspectors identified a noncited violation of License Condition 2.C(41) for the failure to identify conditions adverse to the fire protection program. Specifically, during required inspections of the material condition of the sprinkler system, the licensee failed to identify several instances of bent or misaligned sprinkler head deflector plates and a painted sprinkler head. Corrective action included correcting bent or misaligned plates and replacing the painted sprinkler head. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-03132. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the safety concern is that the number of bent or misaligned sprinkler heat canopies and painted sprinkler heads would not provide an adequate area-wide coverage of suppression. The inspectors evaluated the significance of this finding using Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." The deficiency involved the Fixed Fire Protection Systems category. Using Appendix F, Enclosure , "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," the inspectors determined that the deficiency had low degradation since less than 10 percent of the heads in the affected fire area were nonfunctional, a functional head remained within 10 feet of the combustibles of concern, and the system remained nominally code compliant. This finding screened as having very low safety significance (Green) in Phase 1of Manual Chapter 0609, Appendix F. This finding has a cross-cutting aspect in the area of human performance associated with resources because the procedure used to inspect the condition of these sprinklers did not contain specific criteria for identifying unacceptable sprinkler conditions [H.2(c)](Section 1R05). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure that Safety Related Manholes were Properly Sealed to Prevent the Entry of Flammable Liquid Green. The inspectors identified a noncited violation of Facility Operating License Condition 2.C(41), involving the failure to ensure that manholes MH01, MH20 and MH21 were properly sealed to prevent the entry of flammable liquid. During the performance of the manhole/vault inspection, the inspectors were reviewing engineering change packages associated with solar sump pumps for MH20 and MH21. During their review, they determined that the licensee was not meeting the requirements of their license bases documents for MH20 and MH21, which contain safe shutdown cables for standby service water trains A and B. The licensees immediate corrective action included placing hazmat barricades around each manhole to prevent flammable fluids from entering the manholes. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-00562. This finding was more than minor because it was associated with the protection against external factors attribute of the

Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 3b, Item 1 directs the inspectors to Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. However, an NRC senior reactor analyst determined that the unique nature of this performance deficiency did not lend itself to analysis by the methods provided in Appendix F. Therefore, a Phase 3 analysis was performed. Based on a bounding analysis, the analyst determined that the change in core damage frequency was approximately 1.5E-7/yr. The result was low because of the relatively short periods of time that fuel was actually being transferred, the low probability of transfer system failures, and the low likelihood that a loss of normal service water initiator would occur following a fire in the subject manholes. This noncited violation was therefore determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective actions because licensee personnel failed to initiate a condition report when the issue was identified during the development of their engineering change package, which resulted in the failure to ensure the safety related manholes were sealed in accordance with their license based documents [P.1(a)](Section 1R06). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Provide Adequate Procedures for High Pressure Core Spray Minimum Flow Valve Surveillance Testing Green. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a for the licensees failure to provide adequate testing procedures, which resulted in the high pressure core spray minimum flow valve inadvertently stroking approximately 11 times during a surveillance test. The excessive stroking of the valve resulted in the unplanned inoperability of the high pressure core spray system because the valves feeder breaker overcurrent instantaneous trip setpoint had drifted below the manufacturers tolerance for the existing setting. As immediate corrective action, the licensee replaced the degraded breaker. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01901. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone's objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function since the high pressure core spray system would still have been functional even with the minimum flow valve potentially failing open. Additionally, it did not represent a loss of a system safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with operating experience in that licensee had not incorporated operating experience from a similar event that had occurred at another Entergy site [P.2(b)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Loose Fuse Clips in Division 3 Emergency Diesel Generator Green. The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take adequate corrective actions for a significant condition adverse to quality associated with the division 3 emergency diesel generator. While performing a maintenance effectiveness review of the diesel generators, the inspectors noted on October 17, 2009, at 9:07 p.m., the FU-7 fuse for the division 3 diesel generator was determined to have a faulty fuse clip, resulting in the inoperability of the diesel generator due to loss of power to the direct current powered fuel pumps. Then on March 18, 2011, the division 3 emergency diesel generator was again rendered inoperable due to a faulty fuse clip on the FU-8 fuse holder, which is of the same design and function as the FU-7 fuse holder in the previous occurrence. Short term corrective action included replacing the fuse holder. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01868.

The finding is more than minor because it is associated with equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, 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 in the area of human performance associated with resources because the training provided to correct the initial event was not adequate to ensure proper fuse installation and verify good connection existed between the fuse and fuse holder [H.2(b)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Assure Configuration Control of Safety Related Systems Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to review the suitability of leaving test fittings on reactor coolant system flow transmitter equalizing block drain ports instead of the design specified manifold plugs. As corrective action, the licensee replaced the test fittings with the correct drain plugs. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-04485. This finding is more than minor because it is associated with the design control attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability of functionality, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, associated with work practices, because the licensee failed to ensure that human error prevention techniques, such as holding pre-job briefings, self- and peer-checking, and proper documentation of activities were utilized such that work activities were performed safely and personnel did not proceed in the face of uncertainty or unexpected circumstances Specifically, the licensee failed to review the suitability of installing test and brass fittings on pressure, differential pressure and flow transmitter block valve drain ports instead of the design specified manifold plugs. [H.4(a)](Section 1R12). Inspection Report# : 2011003 (pdf) Significance: Jun 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow a Procedure Resulting in the Inoperability of the Reactor Core Isolation Cooling System Primary Containment Isolation Valve Green. The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, for failure to follow a procedure resulting in the inoperability of the reactor core isolation cooling system primary containment isolation valve. This occurred while the licensee was performing surveillance on the reactor core isolation cooling system and incorrectly attached a jumper to the wrong terminal point resulting in blowing a fuse that caused a loss of control power to the reactor core isolation cooling primary containment isolation valve 1E51-F031. As immediate corrective action, the licensee removed the jumper and replaced the control power fuse. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2011-01932. The finding is more than minor since it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of a system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather

initiating event. In addition, this finding had a human performance cross-cutting aspect associated with work practices in that the licensee did not use the proper human performance techniques of self-checking to prevent the loss of control power to a primary containment isolation valve [H.4(a)](Section 1R22). Inspection Report# : 2011003 (pdf) Barrier Integrity Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Modification of the Spent Fuel Pool without Prior NRC Approval SLIV. The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests and Experiments, when the licensee failed to obtain a license amendment prior to implementing a proposed change to the plant that required a change to Technical Specifications. The 10 CFR 50.59 evaluation performed by the licensee is dated January 24, 2001, thus it was performed under the requirements of the old rule based on the Entergy Operations letter dated March 5, 2001. In the 10 CFR 50.59 evaluation for the removal of Blackness Testing and the division of the spent fuel pool into two regions, the licensee determined that the modifications did not require a change to Technical Specifications. However, 10 CFR 50.36, Technical Specifications, Section 4, Design Features, requires that design features such as geometric arrangements, which, if altered or modified, would have a significant effect on safety, must be incorporated into Technical Specifications. The NRC considers that the establishment of two regional zones in the spent fuel pool, each having specific loading criteria to maintain keff less than 0.95, constitutes design features which, if altered or modified would have a significant effect on safety. Therefore, these design features should have been incorporated into the Technical Specifications. In a letter dated September 8, 2010, (ML102660403), the licensee submitted a power up-rate license amendment request. The NRC staff is currently reviewing the license request, which includes the licensees technical justification for the spent fuel pool changes described above. Based on preliminary review of the amendment request, the NRC staff has determined that an immediate safety concern does not exist. The licensee has entered this issue into their corrective action program as condition report CR-GGN-2012-01077. The finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, containment) protect the public from radionuclide releases caused by accidents or events. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characteristics of Findings, and determined that the finding was of very low safety significance (Green) because it did not result in the loss of cooling to the spent fuel pool, did not result from fuel handling errors that caused damage to fuel clad integrity, and it did not result in a loss of spent fuel pool inventory. This finding is a latent issue and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 1R15). Inspection Report# : 2012002 (pdf) Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Ensure Correct Fuses were Installed in the Hydrogen Igniter Control Circuits Green. The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to install the correct amperage fuses in the hydrogen igniter control circuit. On August 4, 2011, the inspectors were performing an operability review of a condition report dealing with the division 1 hydrogen igniters. The licensee had determined that half the division 1 hydrogen igniters would not energize, and in their investigation, they determined that the loss of power to the hydrogen igniters was due to a blown fuse. The licensee also determined that the blown fuse was 0.3 amps and should have been 0.8 amps per plant drawings. The licensee performed an operability determination for the as found condition and determined that the circuit required 0.193 amps to power the circuit, which included the light bulbs. The inspectors reviewed the operability determination and the calculations and determined that the licensees conclusions were reasonable. The licensee immediate corrective action was to replace the incorrect fuses one division at a time with the correct size 0.8 amp fuses and

restore the hydrogen igniters to operable status. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-005388. This finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone, and it adversely affected the cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that Appendix H, Containment Integrity Significance Determination Process, was required. Inspectors determined that this was a type B finding and, using section 6.0, determined that the finding was of very low safety significance (Green) because during their review, the inspectors noted that the hydrogen igniters had maintain functionality over the life of the plant based on satisfactory surveillance tests and no previous failures. Therefore, the exposed time for the de-energized hydrogen igniters was less than 3 days, resulting in very low safety significance. The Appendix H evaluation and the final risk significance determination were reviewed and concurred on by a regional senior reactor analyst. This issue is a latent issue associated with original plant equipment and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 1R15). Inspection Report# : 2011004 (pdf) Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform Preventative Maintenance on the Fuel Handling Bridge Paddle Switch Green. The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for inadequate preventative maintenance instructions, which resulted in the loss of control of the fuel handling bridge in the spent fuel pool. On July 15, 2011, while moving spent fuel from the spent fuel pool to the dry cask loading pool, the fuel handling platform did not stop when the paddle switch was released from the reverse position. The paddle switch did not return to the neutral position as designed, and the bridge continued to move in the reverse direction. The fuel handling bridge tripped the zone limit switches and came to a stop. The licensee concluded that the switches had to be cleaned, adjusted, and re-greased periodically to ensure proper operation. Immediate corrective actions included replacing the paddle switch and revising the preventive maintenance instruction to clean and re-grease the paddle switch before every dry cask fuel campaign. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2011-04896. The finding is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases caused by accidents or events. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not result from fuel handling errors that caused damage to fuel clad integrity because the fuel handling bridge movement was arrested prior to coming in contact with the spent fuel pool wall. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with the operational experience component because the licensee failed to evaluate the need to update the preventative maintenance procedure for known issues associated with the fuel handling bridge paddle switch prior to the implementation of the dry fuel storage campaign [P.2(b)] (Section 4OA2). Inspection Report# : 2011004 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety

Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: N/A Dec 01, 2011 Identified By: NRC Item Type: FIN Finding Grand Gulf, 2011, Biennial Problem Identification and Resolution Inspection Summary The inspectors concluded that the licensee was, in general, effective in identifying, evaluating, and resolving problems. Grand Gulf personnel were identifying and entering issues into the corrective action program at appropriately low thresholds as evidenced by a large number of condition reports issued. The team determined that the licensee generally screened issues appropriately for operability and reportability. The team noted that issues were typically identified promptly and prioritized commensurate with their safety significance. Most root and apparent cause analyses appropriately considered extent of condition and previous occurrences. The team concluded that the corrective actions were generally identified and implemented promptly. The team found that the licensee had established and was maintaining an environment at Grand Gulf where employees felt free to raise safety concerns without fear of retaliation. The licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items in the corrective action program. The licensee used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective quality assurance audits and self assessments, as demonstrated by self identification of corrective action program areas for improvement. Inspection Report# : 2011006 (pdf) Last modified : May 29, 2012

Grand Gulf 1 2Q/2012 Plant Inspection Findings Initiating Events Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of Hot Work Fire Watch to Follow Procedural Requirements The inspectors reviewed a self-revealing non-cited violation of Technical Specifications 5.4.1(a), for failure of the hot-work fire watch to follow procedural requirements, which resulted in a fire in main condenser A. On April 11, 2012, at 6:11 p.m., hot-work was in progress inside the condenser A in the upper southeast corner at 150 foot elevation. Cutting was being performed by contract boilermakers using an oxy-acetylene torch, with ventilation exhaust and supply provided by nearby HEPA hoses. The torch cutting operation produced hot slag, which exited the barrier provided by the fire blankets and ignited the nearby HEPA hoses, air conditioning hoses, and eventually the acetylene hoses. Contract pipefitters in the area were able to extinguish the fire. The main control room was informed of the fire inside condenser A and dispatched the fire brigade to the scene. The operations shift manager declared a notice of unusual event at 6:26 p.m. due to a fire in the protected area lasting longer than 15 minutes. Members of the fire brigade entered the condenser bay at 6:42 p.m. and reported to the control room there was no fire present, only smoke. The notice of unusual event was exited at 7:00 p.m. Short term corrective actions included site management placing a stop work order on all hot-work until a complete investigation of the event could be performed. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-05418. The finding is more than minor because it is associated with the protection against external factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors reviewed Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," that states in the Assumptions and Limitations section, The Fire Protection SDP focuses on risks due to degraded conditions of the fire protection program during full power operation of a nuclear power plant. This tool does not address the potential risk significance of fire protection inspection findings in the context of other modes of plant operation (i.e., low power or shutdown). Therefore, the senior reactor analyst evaluated the finding in accordance with Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for both PWRs and BWRs. The finding did not require a quantitative assessment because adequate mitigating equipment remained available; the finding did not increase the likelihood of a loss of reactor coolant system inventory; the finding did not degrade the ability to terminate a leak path or add reactor coolant system inventory; and the finding did not degrade the ability to recover decay heat removal if lost. Therefore, the finding screened as Green, having very low safety significance. The inspectors determined that the apparent cause of this finding was that site management did not ensure that hot-work supervisors were engaged in ensuring compliance with procedural requirements. This finding had a cross-cutting aspect in the area of human performance associated with work practices component because the licensee failed to ensure supervisory oversight of hot-work activities is performed within procedural requirements such that nuclear safety is supported [H.4(c)] (Section 40A3) Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: Self-Revealing Item Type: FIN Finding Manual Reactor Scram Caused by Failure to Ensure the Main Steam Supply Valve to Reactor Feed Pump Turbine B was Full Open Green . The inspectors reviewed a Green self-revealing finding for the failure to ensure the correct position (full open) of the main steam supply valve 1N11-F014B to reactor feed pump turbine B, which resulted in a manual reactor scram due to decreasing reactor water level. During plant shutdown activities to begin refueling outage 18, the at-the-controls

operator manually scrammed the reactor from approximately 23 percent rated thermal power due to the decreasing reactor water level. Water level in the reactor was decreasing because valve 1N11-F014B was not fully open, and because pressure in the main steam lines had been reduced when the crew opened turbine bypass valves to begin cooling the main turbine. With valve 1N11-F014B less than fully open and reduced steam pressure, the operating feed pump wasnt able to maintain water level. After the scram, reactor core isolation cooling and reactor feed pump turbine A were used to restore water level. The licensee plans to repair valve 1N11-F014B during the current refuelling outage. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01838. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors concluded that the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment would not be available. The inspectors, in consultation with the regional senior reactor analyst, performed a Phase 2 estimation using the pre-solved work sheets for Grand Gulf Nuclear Station. The inspectors determined by entering the power conversion system column that the finding was of very low safety significance (Green). This result was validated by the senior reactor analyst using the current revision of the plant-specific SPAR model. The inspectors determined the finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the operating staff proceeded with the start up of the reactor feed pump B with the main steam supply valve 1N11-F014B in an unknown position [H.1(b)](Section 1R11). Inspection Report# : 2012002 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Online Risk Assessment Per Severe Weather Off Normal Procedure Due to a Declared Tornado Warning Affecting Grand Gulf Nuclear Station Green. The inspectors identified a Green non-cited violation of Technical Specifications 5.4.1.a for the failure to perform an online risk assessment per severe weather off normal procedure due to a declared tornado warning affecting Grand Gulf Nuclear Station. At 7:41 p.m., on February 15, 2012, the National Weather Service issued a tornado warning for Claiborne County, the county in which Grand Gulf Nuclear Station is located. In response to a tornado warning, licensee procedures required them to enter Off-Normal Operating Procedure 05-1-02-VI-2, Severe Weather, and evaluate online risk. This severe weather condition would have resulted in the licensee entering into an orange risk condition. On February 16, 2012, the inspectors identified that the licensee had not made a log entry for entry into their off normal severe weather procedure during the preceding evening and therefore had not evaluated online risk status for the severe weather condition. In response to the inspectors observations, the licensee initiated a condition report detailing the failures to enter the off normal procedure and enter the correct risk condition. The licensee has implemented short-term corrective actions to ensure the site adequately evaluates the risk associated with adverse weather. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01707. The finding is more-than-minor because it is associated with the Initiating Events Cornerstone attribute of protection against external events, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Appendix K; Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1; Assessment of Risk Deficit; and consulting with the regional senior reactor analyst, the inspectors determined the finding to be of very low safety significance based on a licensees calculated determination of the incremental core damage probability deficit of 4.0E-08. This result was validated by the senior reactor analyst using the current revision of the plant-specific SPAR model. The inspectors determined the finding has a cross-cutting aspect in the area of human performance associated with the resources component because the on-shift senior reactor operators did not have adequate access to current weather information that would prompt control room personnel to re-evaluate risk due to changing weather conditions [H.2(d)](Section 1R13). Inspection Report# : 2012002 (pdf)

Mitigating Systems Significance: Jun 30, 2012 Identified By: NRC Item Type: FIN Finding Failure to Ensure Materials are Stored Properly in the 500 KV Switchyard The inspectors identified a finding for the licensees failure to ensure that materials or equipment were not stored under energized lines or near energized equipment in accordance with station procedures. On May 21, 2012, the inspectors were performing a grid stability inspection and toured the 500 KV switchyard with the system switchyard engineer. During the tour, the inspectors identified numerous cylindrical shaped items stored under a 500 KV power line, which posed a missile hazard to the offsite source of power. The licensee determined that the items in question were bushing sleeves that were left in the switchyard following 500 KV breaker maintenance. The inspectors researched station procedures and determined that the cylindrical items stored under the energized 500 KV power line did not meet procedure requirements for the storage of materials and equipment. Immediate corrective actions included having the items removed from the switchyard. The licensee entered this issue into their corrective action program as Condition Report CR-GGNS-2012-07362. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigation Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors reviewed Manual Chapter 0609, Attachment A, Phase 1 - Initial Screening and Characterization of Findings. Attachment A, Table 4.a, states that a Phase 3 is required if the finding is potentially risk significant due to external initiating event core damage accident sequences. The inspectors determined that the failure to properly store the bushing sleeves in the switchyard could have resulted in a loss of offsite power during a severe weather initiating event. Therefore, the senior reactor analyst evaluated the finding to determine its significance using hand calculations and the site-specific SPAR model. The analyst determined that the probability of having straight-line winds or winds generated by hurricanes or tornados that were strong enough to throw the bushing sleeves into switchyard electrical equipment was between 2.5 x 10-1 and 2.0 x 10-2 /year. The analyst also determined that the conditional probability that bushing sleeves thrown by winds would result in a loss of offsite power was between 1.2 x 10-1 and 1.1 x 10-7. Finally, the SPAR model calculated that the conditional core damage probability for a loss of offsite power initiated in the switchyard was 5.3 x 10-5. Using these values, under all scenarios evaluated by the analyst, the change in core damage frequency caused by the subject performance deficiency was below 1 x 10-6. Therefore, the finding was of very low safety significance (Green). The inspectors determined the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement the corrective action program with a low threshold for identifying materials improperly stored in the 500 KV switchyard [P.1(a)](Section 1R01). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Loss of Alternate Method of Decay Heat Removal Due to Reactor Water Clean Up Pumps Tripping on Low Suction Flow Signal Green. The inspectors reviewed a self-revealing finding for the licensees failure to identify that de-energizing non-safety electrical bus 13BD1 and 13BD2 would cause the reactor water clean-up pumps A and B to trip on a low suction flow signal. On April 24, 2012, the plant was shut down for refueling outage 18, the residual heat removal system B was in service, and the reactor water clean-up system was in standby mode as the alternate shutdown cooling system. In this configuration, the plant was in yellow risk due to having two available systems for decay heat removal. At 10:00 a.m., both reactor water clean-up pumps tripped on low pump suction flow, causing the plant to enter an unplanned orange risk configuration for only having one system available for decay heat removal. The licensee determined the reactor water pumps tripped while opening the feeder breaker for the 13BD1 and 13BD2 buses (breaker 152-1305) for scheduled maintenance. When breaker 152-1305 was opened, optical isolator AT12 caused the pump low suction flow trip control contacts to close, which initiated the low suction flow alarm and caused

the pumps to trip. Immediate corrective actions included restoring reactor water clean-up as the alternative source of decay heat removal by closing breaker 152-1305 and re-energizing the 13BD1 and 13BD2 buses. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2012-06092 and CR-GGN-2012-06105. The finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and it affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a loss of a system safety function. The inspectors determined that the cause of this finding was a latent issue; therefore no cross-cutting aspect was assigned (Section 1R13). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions to Address Configuration Control of Previous Non-cited Violation Green. The inspectors identified a non-cited violation of very low safety significance of 10 CFR Appendix B, Criterion XVI, "Corrective Action," for failure to implement adequate corrective actions for a previous NRC-identified non-cited violation. The previous finding involved a failure to maintain configuration control of various systems in the plant. In response to the previous finding, the licensee performed an apparent cause evaluation and developed actions to address the causes and extent of condition. However, the inspector identified that the actions pertaining to the extent of condition were not properly implemented and, as a result, the deficiency identified by the inspector was not fully resolved. The licensee failed to identify brass compression fittings installed on drain tailpieces of the standby service water system instead of stainless steel fittings as required by design documents. Furthermore, the licensee failed to update applicable design drawings allowing sacrificial compression fittings to be installed. The licensee performed corrective actions to restore configuration control. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-04003, CR-GGN-2012-4180, and CR-GGN-2012-04233. The issue is more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, the issues identified by the inspector impacted the licensees ability to establish and maintain configuration control for equipment relied on for safe operation of the plant. The design control attribute of the Mitigating Systems Cornerstone and the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences were affected. Until the issues are fully resolved, the licensee continues to be vulnerable to gaps in their system configuration control. The finding was determined to be of very low safety significance (Green) using Attachment 4 to IMC 0609, "Significance Determination Process," because it did not result in an actual loss of safety function. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance associated with the resources component because the licensee did not provide adequate training of personnel so that the inappropriately installed fittings could be identified during system walkdowns [H.2(b)] (Section 1R08). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement a Surveillance Requirement to Assure that the Limiting Condition for Operation Will be Met Green. The inspectors identified a non-cited violation of 10 CFR Part 50.36, Technical Specifications, involving the failure to implement a surveillance requirement to assure that the limiting conditions for operation of the ultimate heat sink will be met. Technical Specifications requires two cooling towers and two cooling basins, with the volume of the two basins constituting the entire inventory of the ultimate heat sink. Therefore, an interconnecting siphon line is installed to transfer water between the two cooling tower basins. That siphon line has the safety-related function of ensuring the availability of enough cooling water to satisfy ultimate heat sink requirements. Technical Specification 3.7.1 includes Surveillance Requirement 3.7.1.1, which verifies the water level in each cooling tower basin every 24

hours, and Surveillance Requirement 3.7.1.2, which verifies each cooling tower fan every 31 days. However, the inspectors identified that Technical Specification 3.7.1 does not include a surveillance requirement to verify that the interconnecting siphon line will perform its safety-related function. On May 20, 2012, the licensee performed an operability test for the siphon line and determined that it was operable. The licensee is currently performing a preventative maintenance task as a compensatory action to ensure operability of the siphon line until a license amendment can be submitted to the NRC that establishes a surveillance requirement. The licensee documented this violation in Condition Reports CR-GGN-2012-08257 and CR-GGN-2012-08537. The violation is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, without a surveillance requirement that verifies the interconnecting siphon line can perform its safety-related function, the licensee cannot ensure that sufficient cooling water is available following an accident. The inspectors evaluated the finding using Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings and determined that the finding was of very low safety significance (Green) because the finding was a design or qualification deficiency confirmed not to result in a loss of operability or function; did not represent a loss of safety system function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time; 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 in the human performance area associated with the resources component because the licensee did not ensure that equipment was adequate to assure nuclear safety, in that the licensee had recently reviewed documentation associated with a modification to the siphon line but failed to identify that operability of the UHS could not be established without a technical specification surveillance requirement to ensure operability of the siphon line [H.2(c)] (Section 1R19). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow a Post-Modification Test Procedure Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to follow a post-modification test procedure for the interconnecting siphon line between the two standby service water system cooling tower basins. Operability of the ultimate heat sink is based on a minimum water level in the two standby service water cooling tower basins, an operable interconnecting siphon between the basins, and four operable cooling tower fans (two per basin). At extended power uprate conditions, the configuration of the basins and the original siphon line would not support 30 days of operation of both trains of the standby service water system and the high pressure core spray service water systems without makeup, so the licensee performed a modification (EC 25649), which involved replacing the original siphon line with a new siphon line in order to transfer water from one basin to the other. On March 28, 2012, after completing the modification, the licensee performed post-modification testing to determine the piping friction loss coefficient of the modified siphon line and to evaluate its acceptability against the worst-case friction loss coefficient documented in EC 25649. The licensee deviated from the test procedure, as-written, and performed the test with an inadequate pressure gauge instead of the specified gauge. After inspectors challenged the validity of these test results, the licensee performed another test of the siphon line with a different method that did not require the use of a pressure gauge to measure the piping friction loss coefficient. The inspectors reviewed the subsequent test data and found the test results to be satisfactory. The licensee documented this concern in Condition Report CR-GGN-2012-05260. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the use of an unqualified gauge invalidated the test results, and a different test method had to be developed to determine the piping friction loss coefficient for the siphon line. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed to result in loss of operability or function; did not represent a loss of safety system function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time;

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 in the human performance area associated with work practices component because licensee personnel proceeded in the face of uncertainty or unexpected circumstances. Specifically, the licensee proceeded with the test without verifying that the pressure gauge was suitable for the test conditions after observing unexpected measurements with the gauge [H.1(a)] (Section 1R19). Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions Associated with Division 1 and 2 Standby Service Water Safety Related Cables that were Partially Submerged in Cable Manhole/Vault Green . The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take timely corrective actions to correct a condition adverse to quality associated with division 1 and 2 standby service water safety related cables that were partially submerged in a cable manhole/vault. The inspectors reviewed work order 52284535 and noted that the sump pump for manhole/vault MH-01, which contained standby service water cables for division 1 and 2, was determined to be non-functional on September 10, 2011. The inspectors determined that a work order to repair the non-functioning sump pump had been developed but that the work order had not yet been scheduled. During a subsequent inspection, manhole/vault MH-01 was found to contain approximately three feet of water, with water partially covering some of the safety related cables. The electricians immediately pumped manhole/vault MH-01 and wrote a condition report. The licensee repaired the sump pump the next week and declared it functional. The cables remained operable based on the results of meggar tests. The licensee entered this issue into their corrective action program as condition reports CR-GGN-2012-00503, 01324, and 01389. The finding is more than minor because it is associated with the equipment performance attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent either a loss of system safety function or an actual loss of safety function of a single train of one or more non-Technical Specification 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 in the area of human performance associated with the work practices component, in that the licensee personnel did not initiate a condition report as required by licensee procedure when the work order associated with sump pump testing of MH-01 determined that the sump pump was not functioning properly [H.4(b)] (Section 1R06). Inspection Report# : 2012002 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Condition Adverse to Fire Protection, in That the Licensee Failed to Adequately Provide Contingency Lighting in the Fire Brigade Dress Out Area While Normal Lighting was Inoperabl Green. The inspectors identified a Green non-cited violation of Facility Operating License Condition 2.C(41), for the failure to correct a condition adverse to fire protection. Specifically, the licensee failed to adequately provide contingency lighting in the fire brigade dress out area while normal lighting was inoperable due to maintenance on an associated breaker. The inadequate lighting delayed fire brigade response to a potential fire in the turbine building. Immediate corrective action included placing temporary lighting in the area. Normal lighting to the area was restored the next week. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01488. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined

from table 3b that issues related to performance of the fire brigade are not included in Appendix F and require NRC management review using Appendix M. Regional management review evaluated the overall impact of lighting issue in the fire brigade dress out area and concluded that, while the fire protection defense-in-depth was affected by the performance deficiency, the overall defense-in-depth of the front-line systems was not impacted because of train separation and safe shutdown analysis at the site. Therefore the finding screened as having very low safety significance (Green) in accordance with Manual Chapter 0609, Appendix M. The inspectors determined the finding had a cross-cutting aspect in the area of human performance associated with the work control component, in that licensee personnel failed to ensure adequate job site conditions (lighting in the fire bridge dress out area) were in place prior to performance electrical maintenance in the turbine building [H.3(a)] (Section 40A3). Inspection Report# : 2012002 (pdf) Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation Door Seals Protecting Safety Related Equipment Green. The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. Inspectors found that one of the door seals to standby service water pump house A was in a degraded condition. The inspectors identified that the door seal did not make complete contact with the door frame all the way around. The licensee determined that the probable maximum precipitation seal for the identified door was in a degraded condition. Failure of this door seal during a probable maximum precipitation event could potentially cause flooding of the standby service water pump house A. Immediate corrective actions included the site initiating compensatory actions for the degraded seal by staging sand bags in the area and requiring monitoring of the affected door during heavy rainfall. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2011-07687. The finding is more than minor because it is associated with the protection against external factors attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors used the seismic, flooding, and severe weather Table 4b and determined that it would not affect multiple trains of safety equipment and that the finding had very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the resources component in that the licensees procedure used for the inspection of the door seals did not take into account the status of the pump house ventilation system while performing the door seal inspection, and therefore, the licensee failed to make the required adjustments to the door seals resulting in their inspections of the probable maximum precipitation door seals being inadequate [H.2(c)] (Section 1R05). Inspection Report# : 2011005 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions for Reactor Core Isolation Cooling System Venting Green. The team identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct an inadequate venting procedure for the reactor core isolation cooling system. Corrective actions were not taken in a timely enough manner such that resolution was reached prior to time to demonstrate the licensee met their applicable technical specification surveillance requirement. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-07669 and subsequently altered their procedure, which performs the technical specification surveillance requirement to demonstrate that it meets the applicable requirements. This finding is more than minor because it affects the procedure quality 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. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of

a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the decision making component. The licensee did not use conservative assumptions when deciding to pursue corrective action for venting of the reactor core isolation cooling system piping to demonstrate their action was safe in order to proceed rather than demonstrating it was unsafe to disapprove the action [H.1(b)]. (Section 4OA2.5a) Inspection Report# : 2011006 (pdf) Significance: SL-IV Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit a Licensee Event Report for an Inoperable Reactor Core Isolation Cooling System Severity Level IV. The team identified a Severity Level IV noncited violation of 10 CFR 50.73, Licensee Event Report System, associated with the licensees failure to submit a licensee event report within 60 days following discovery of an event meeting the reporting criteria as specified. Specifically, the licensee was not meeting the technical specification surveillance requirement for venting the reactor core cooling isolation system and subsequently the system was inoperable in excess of the allowed outage time which constituted a condition prohibited by technical specifications. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8890. This finding affects the mitigating systems cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRCs ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section 6.9 of the Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect, as it is not indicative of current performance (Section 4OA2.5b). Inspection Report# : 2011006 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Document a Condition as a Significant Condition Adverse to Quality Green. The team identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to identify and document a significant condition adverse to quality and report the condition to appropriate levels of management. As a result, a root cause analysis was not performed and more comprehensive actions to prevent recurrence were not considered for the condition. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011- 07671, to address the problem. This finding is more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance (Green) because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensees procedures for significant conditions adverse to quality were not complete and accurate enough to prevent the condition. [H.2(c)]. (Section 4OA2.5c) Inspection Report# : 2011006 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: VIO Violation

Inadequate Corrective Action for a Leak on the Division II Emergency Diesel Generator Lube Oil Sump Green. The team identified a Green cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a leak on the Division II emergency diesel generator lube oil sump. Despite identification of the leak in 2004, ineffective attempts to repair the leak and previous identification by the NRC in 2009, the licensee dispositioned the leak as accept as-is without a full understanding of the lube oil sump leak and potential consequences. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8880. The condition was discovered and documented by the licensee in 2004. This finding was initially determined by the NRC to be a minor violation in 2009. Paragraph F of Section 2.10 of the NRC Enforcement Manual states in part that where a licensee does not take corrective action for a minor violation, the matter should be considered more than minor and associated with a green inspection finding and dispositioned in a cited or noncited violation, as appropriate. This finding is now determined to be more than minor because if left uncorrected the failure to restore the lube oil sump for the Division II emergency diesel generator to design conditions would have the potential to lead to a more significant safety concern, specifically, the leak could worsen and potentially affect operability of the emergency diesel generator. Due to the licensees failure to restore compliance within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy. This finding affects the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate this problem such that the resolutions addressed the causes [P.1(c)]. (Section 4OA2.5d) Inspection Report# : 2011006 (pdf) Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Establish Preventative Maintenance for Components Used in Critical Applications Green. The inspectors reviewed a self-revealing, noncited violation of Technical Specification 5.4.1.a for the licensees failure to follow a procedure that required them to evaluate components of critical systems in order to establish a preventive maintenance strategy, which resulted in unscheduled unavailability of safety-related systems and associated unscheduled entries into 72-hour shutdown Technical Specification action statements. The inspectors noted the following two examples dealing with failures of safety related equipment, which resulted in entering into shutdown limiting condition of operation. On June 2, 2011, Grand Gulf Nuclear Station experienced a failure of a relay in the standby service water B pump house ventilation system, which rendered the standby service water B system inoperable. The immediate corrective actions were to replace the relays and to restore the ventilation system. On June 22, 2011, the station experienced a failure of a motor contactor coil on breaker 52-154128, which caused the engineered safety feature electrical switchgear room cooler fan coil unit 1T46B003A not to run. The maintenance personnel determined the failure was due to a burnt motor contactor coil. The immediate corrective action was to replace the contactor coil and restore the room cooler. In both cases, the failed equipment was original plant equipment and preventive maintenance measures had not been established. The licensee entered these issues into the corrective action program as Condition Reports CR-GGN-2011-3730 and CR-GGN-2011-4313. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not result in a loss of system safety function of a single train for more than its technical specifications allowed outage time. This issue is a latent issue associated with original plant equipment and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 4OA2). Inspection Report# : 2011004 (pdf)

Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Provide An Adequate Alternative Shutdown Procedure

  • Green. The team identified a noncited violation of License Condition 2.C(41), Fire Protection Program, for failing to ensure that the postfire safe shutdown procedure for fires requiring control room evacuation could be performed within the critical times required by the approved fire protection program. Specifically, two crews of operators simulating performance of Procedure 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Revision 036, did not give priority to the required safe shutdown components which are protected against fire damage and did not complete the equipment alignments within the times required by the thermal-hydraulic analysis. The team confirmed at the end of each walkdown that the operators involved did not know what the credited shutdown equipment was for a postfire safe shutdown or the critical time limits to be met. The team also confirmed that the licensee had not performed timed walkdowns to validate that the procedure would complete the required actions for postfire safe shutdown within the times required by the thermal-hydraulic analysis. The licensee entered this into their corrective action program as CR GGN 2011 02721, implemented compensatory measures to focus the operators priority on the required safe shutdown components and implemented a procedure revision.

The failure to provide an adequate procedure to implement the requirements of the approved fire protection program for a fire in the control room is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. The scenario impacted operators being ready to emergency depressurize the reactor and reflood using a residual heat removal pump. Because a bounding change to core damage frequency was 4.13 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding is of very low safety significance (Green). The finding did not have a crosscutting aspect since the primary cause did not fit any crosscutting aspects. (Section 1R5.5.b.1) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Assure Equipment Required For Postfire Safe Shutdown Was Protected Against Fire Damage Green. The team identified a noncited violation of License Condition 2.C(41), Fire Protection Program, for failing to assure that equipment relied upon for safe shutdown following a fire in the control room was protected against fire damage. Specifically, Procedure 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, Revision 036, relied on the automatic operation of and indications from the load shedding and sequencing system. The team identified that this system was not isolated from potential damage due to a fire in the control room and the procedure did not adequately address the potential that fire damage to the system could effect the postfire safe shutdown capability by spuriously starting or stopping electric loads. The licensee entered this into their corrective action program as CR GGN 2011 02721. The failure to assure that equipment required to successfully implement the safe shutdown procedure for a fire in the control room was protected against fire damage is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 1.97 x 10-8, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding did not have a crosscutting aspect since it was not indicative of current performance, in that the licensee had

established the current procedure more than three years prior to this finding. (Section 1R5.5.b.2) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions to Protect Safe Shutdown Equipment From Fire Damage Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to take timely corrective action to modify the control circuits for 33 motor operated valves that are relied upon during safe shutdown due to fire. Noncited violation NCV 05000416/2008006-04, Failure to Ensure That Damage to Motor-Operated Valve Circuits Would Not Prevent Safe Shutdown, documented the licensees inadequate review of Information Notice 92-18, Potential for Loss of Remote Shutdown Capability During Control Room Fire. The licensee failed to develop modification packages such that motor operated valve control circuit modifications could be implemented during the fall 2010 refueling outage. As a result, 33 motor operated valves associated with safe shutdown equipment continue to remain susceptible to potential damage during spurious operation due to circuit hot shorts. The licensee has maintained a fire watch as a compensatory measure. The licensee entered this into their corrective action program as CR GGN 2011 02779. The failure to take timely corrective actions to address the potential for fire induced hot shorts to impact the ability to safely shutdown the plant following a fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 9.58 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding had a crosscutting aspect in the area of Human Performance associated with Decision Making, because the licensee failed to demonstrate that nuclear safety is an overriding priority. Specifically, the licensee did not promptly initiate control circuit reviews and implement modifications required for corrective actions after the licensees inadequate evaluation of Information Notice 92-18 was identified in the 2008 violation. [H.1(a)] (Section 1R5.6) Inspection Report# : 2011007 (pdf) Significance: Aug 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions To Assure Postfire Safe Shutdown Green. The team identified a noncited violation of 10 CFR 50, Appendix B, Criterion XVI, for inadequate corrective actions to address the potential for fire induced hot shorts to impact the ability to trip a control rod group as described in Information Notice 2007-07. The licensees evaluation of Information Notice 2007-07 stated in part, provisions have been included in 05-1-02-II-1, Shutdown from the Remote Shutdown Panel, to trip the proper reactor protective system breakers to ensure that the reactor scram occurs. The team noted that Procedure 05-1-02-II-1 contained a conditional statement for the operator to determine if opening the reactor protective system breakers is required. The procedure did not provide assurance that all control rod groups insert since the control room indications to be utilized by the operator were not identified and confirmed to be reliable during fires requiring control room evacuation. The licensee entered this finding into its corrective action program under CR-GGN-2011-02780, implemented compensatory measures to ensure the operators de-energized the reactor protection system, and implemented a procedure change. The failure to take adequate corrective actions to address the potential for fire induced hot shorts to impact the ability to safely shutdown the plant following a fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the reactor safety mitigating systems cornerstone attribute for protection against

external events (fire), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved control room evacuation, a Phase 3 SDP risk assessment was performed by a senior reactor analyst. Because a bounding change to core damage frequency was 9.58 x 10-7, and the finding was not significant with respect to large, early release frequency, this finding was determined to have very low safety significance (Green). The finding did not have a crosscutting aspect since it was not indicative of current performance. The licensee had incorrectly assessed the applicability of Information Notice 2007-07 more than three years prior to this finding. (Section 4OA2.b) Inspection Report# : 2011007 (pdf) Barrier Integrity Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Loss of Decay Heat Removal to the Spent Fuel Pool Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specifications 5.4.1(a), involving a loss of decay heat removal in the spent fuel pool due to station personnel failing to correctly follow operation of pool gate seal air supply procedure. On April 17, 2012, Grand Gulf Nuclear Station was preparing to drain the reactor cavity to reinstall the vessel head after the completion of refueling activities. In preparation, the upper containment pool to the reactor cavity gate was installed by General Electric-Hitachi technicians with Entergy oversight. Technicians were directed by procedure to verify that all supply isolation toggle valves to the gate seals were open and secured in place. However, technicians failed to complete this action correctly and the control room was informed that all prerequisites were completed and began the cavity drain down. The control room immediately noticed the fuel pool drain tank level was decreasing and attempted to makeup to the tank via the normal makeup valve. When the fuel pool drain tank level reached 17 percent full, both fuel pool cooling and cleanup pumps tripped as expected, resulting in loss of decay heat removal to the spent fuel pool. The main control room entered the off-normal event procedure for inadequate decay heat removal, and they secured the drain down evolution. Approximately 47 minutes later, spent fuel pool cooling was re-established. During this event, the spent fuel pool temperature did not exceed the limits required by Technical Requirements Manual Section 6.7.4 (140°F). Short term corrective actions included restoring decay heat removal to the spent fuel pool and conducting a human performance review of the event. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-05756. The finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represented a loss of spent fuel pool cooling that would not preclude restoration of cooling to the spent fuel pool prior to pool boiling. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to use adequate self- and peer-checking techniques to ensure gate seals were properly inflated prior to cavity drain down [H.4(a)] (Section 1R20). Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Modification of the Spent Fuel Pool without Prior NRC Approval SLIV. The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests and Experiments, when the licensee failed to obtain a license amendment prior to implementing a proposed change to the plant that required a change to Technical Specifications. The 10 CFR 50.59 evaluation performed by the licensee is dated January 24, 2001, thus it was performed under the requirements of the old rule based on the Entergy Operations

letter dated March 5, 2001. In the 10 CFR 50.59 evaluation for the removal of Blackness Testing and the division of the spent fuel pool into two regions, the licensee determined that the modifications did not require a change to Technical Specifications. However, 10 CFR 50.36, Technical Specifications, Section 4, Design Features, requires that design features such as geometric arrangements, which, if altered or modified, would have a significant effect on safety, must be incorporated into Technical Specifications. The NRC considers that the establishment of two regional zones in the spent fuel pool, each having specific loading criteria to maintain keff less than 0.95, constitutes design features which, if altered or modified would have a significant effect on safety. Therefore, these design features should have been incorporated into the Technical Specifications. In a letter dated September 8, 2010, (ML102660403), the licensee submitted a power up-rate license amendment request. The NRC staff is currently reviewing the license request, which includes the licensees technical justification for the spent fuel pool changes described above. Based on preliminary review of the amendment request, the NRC staff has determined that an immediate safety concern does not exist. The licensee has entered this issue into their corrective action program as condition report CR-GGN-2012-01077. The finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, containment) protect the public from radionuclide releases caused by accidents or events. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characteristics of Findings, and determined that the finding was of very low safety significance (Green) because it did not result in the loss of cooling to the spent fuel pool, did not result from fuel handling errors that caused damage to fuel clad integrity, and it did not result in a loss of spent fuel pool inventory. This finding is a latent issue and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 1R15). Inspection Report# : 2012002 (pdf) Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Ensure Correct Fuses were Installed in the Hydrogen Igniter Control Circuits Green. The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to install the correct amperage fuses in the hydrogen igniter control circuit. On August 4, 2011, the inspectors were performing an operability review of a condition report dealing with the division 1 hydrogen igniters. The licensee had determined that half the division 1 hydrogen igniters would not energize, and in their investigation, they determined that the loss of power to the hydrogen igniters was due to a blown fuse. The licensee also determined that the blown fuse was 0.3 amps and should have been 0.8 amps per plant drawings. The licensee performed an operability determination for the as found condition and determined that the circuit required 0.193 amps to power the circuit, which included the light bulbs. The inspectors reviewed the operability determination and the calculations and determined that the licensees conclusions were reasonable. The licensee immediate corrective action was to replace the incorrect fuses one division at a time with the correct size 0.8 amp fuses and restore the hydrogen igniters to operable status. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-005388. This finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone, and it adversely affected the cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, inspectors determined that Appendix H, Containment Integrity Significance Determination Process, was required. Inspectors determined that this was a type B finding and, using section 6.0, determined that the finding was of very low safety significance (Green) because during their review, the inspectors noted that the hydrogen igniters had maintain functionality over the life of the plant based on satisfactory surveillance tests and no previous failures. Therefore, the exposed time for the de-energized hydrogen igniters was less than 3 days, resulting in very low safety significance. The Appendix H evaluation and the final risk significance determination were reviewed and concurred on by a regional senior reactor analyst. This issue is a latent issue associated with original plant equipment and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 1R15). Inspection Report# : 2011004 (pdf)

Significance: Sep 27, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Perform Preventative Maintenance on the Fuel Handling Bridge Paddle Switch Green. The inspectors reviewed a self-revealing, noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for inadequate preventative maintenance instructions, which resulted in the loss of control of the fuel handling bridge in the spent fuel pool. On July 15, 2011, while moving spent fuel from the spent fuel pool to the dry cask loading pool, the fuel handling platform did not stop when the paddle switch was released from the reverse position. The paddle switch did not return to the neutral position as designed, and the bridge continued to move in the reverse direction. The fuel handling bridge tripped the zone limit switches and came to a stop. The licensee concluded that the switches had to be cleaned, adjusted, and re-greased periodically to ensure proper operation. Immediate corrective actions included replacing the paddle switch and revising the preventive maintenance instruction to clean and re-grease the paddle switch before every dry cask fuel campaign. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2011-04896. The finding is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected cornerstones objective to ensure that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases caused by accidents or events. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not result from fuel handling errors that caused damage to fuel clad integrity because the fuel handling bridge movement was arrested prior to coming in contact with the spent fuel pool wall. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with the operational experience component because the licensee failed to evaluate the need to update the preventative maintenance procedure for known issues associated with the fuel handling bridge paddle switch prior to the implementation of the dry fuel storage campaign [P.2(b)] (Section 4OA2). Inspection Report# : 2011004 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary. Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous

Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Use of Waivers to Allow Workers to Exceed the Minimum Day Off Rule Green. The inspectors identified a non-cited violation of 10 CFR 26, Subpart I, Managing Fatigue, Subsection 207, Waivers and Exceptions, when the licensee inappropriately used waivers to allow workers to exceed the minimum day off rule. While reviewing condition reports, the inspectors noted the use of work hour waivers for a large number of staff. The circumstances for the use of waivers were the refueling outage lasting more than 60 days, contract expiration leading to 14 layoffs, and the loss of 4 workers via voluntary resignation. Due to these circumstances, work hours and fatigue of waivered individuals would have to be assessed daily. The assessment is required because the work hour limit of these individuals exceeded the minimum day off rule, therefore requiring daily monitoring until the end of the cycle. The waivered individuals averaged two days off per six-week period compared to the required three days off. Title 10 CFR 26.207 (a)(2) allows the granting of waivers only to address circumstances that could not have been reasonably controlled. The inspectors determined that the licensee was aware of the circumstances of an extended refueling outage and contract renewal deadline well in advance of the need to grant waivers, and a reasonable amount of time was available for the licensee to develop and execute contingency plans to negate the need to use waivers. Corrective actions included initiating assessments and waivers for exceeding minimum days off requirements for shift personnel for the six-week period ending May 27, 2012, and returning to the normal on-line work schedule in which adequate manpower is available to meet the requirements of the rule. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-7348. The finding is more than minor because it is associated with the access authorization attribute of the Security Cornerstone, and affected the cornerstone objective to provide assurance that the licensees security system and material control and accounting program use a defense in-depth approach and can protect against (1) the design basis threat of radiological sabotage from external and internal threats, and (2) the theft or loss of radiological materials. Using the Inspection Manual Chapter 0609, Appendix E, Baseline Security Significance Determination Process for Power Reactors, Figures 5 and 6, the finding was determined to have very low security significance because the calculated point total did not exceed the threshold value for a Green non-cited violation. The cumulative total for this finding was zero points, which was calculated by factoring the one impact area (vital areas) against Tier III Element 08.02.08, security force work hours, of the access authorization attribute, which resulted in a total of zero points within this attribute. The finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision making component in that the licensee failed to use conservative assumptions in developing staff schedules for the duration of refueling outage 18 and for allowing an employment contract to expire that led to 14 individuals being laid off without realizing the impact these decisions would have on the licensees ability to meet the requirements of the rule [H.1(b)] (Section 1R20). Inspection Report# : 2012003 (pdf) Significance: N/A Dec 01, 2011 Identified By: NRC Item Type: FIN Finding Grand Gulf, 2011, Biennial Problem Identification and Resolution Inspection Summary The inspectors concluded that the licensee was, in general, effective in identifying, evaluating, and resolving problems. Grand Gulf personnel were identifying and entering issues into the corrective action program at appropriately low thresholds as evidenced by a large number of condition reports issued. The team determined that the licensee generally screened issues appropriately for operability and reportability. The team noted that issues were typically identified promptly and prioritized commensurate with their safety significance. Most root and apparent cause analyses appropriately considered extent of condition and previous occurrences. The team concluded that the corrective actions were generally identified and implemented promptly. The team found that the licensee had established and was maintaining an environment at Grand Gulf where employees felt free to raise safety concerns without fear of retaliation. The licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items in the corrective action program. The licensee used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective quality assurance audits and self assessments, as demonstrated by self identification of corrective action program areas for improvement. Inspection Report# : 2011006 (pdf)

Last modified : September 12, 2012 3Q/2012 Inspection Findings - Grand Gulf 1 Grand Gulf 1 3Q/2012 Plant Inspection Findings Initiating Events Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of Hot Work Fire Watch to Follow Procedural Requirements The inspectors reviewed a self-revealing non-cited violation of Technical Specifications 5.4.1(a), for failure of the hot-work fire watch to follow procedural requirements, which resulted in a fire in main condenser A. On April 11, 2012, at 6:11 p.m., hot-work was in progress inside the condenser A in the upper southeast corner at 150 foot elevation. Cutting was being performed by contract boilermakers using an oxy-acetylene torch, with ventilation exhaust and supply provided by nearby HEPA hoses. The torch cutting operation produced hot slag, which exited the barrier provided by the fire blankets and ignited the nearby HEPA hoses, air conditioning hoses, and eventually the acetylene hoses. Contract pipefitters in the area were able to extinguish the fire. The main control room was informed of the fire inside condenser A and dispatched the fire brigade to the scene. The operations shift manager declared a notice of unusual event at 6:26 p.m. due to a fire in the protected area lasting longer than 15 minutes. Members of the fire brigade entered the condenser bay at 6:42 p.m. and reported to the control room there was no fire present, only smoke. The notice of unusual event was exited at 7:00 p.m. Short term corrective actions included site management placing a stop work order on all hot-work until a complete investigation of the event could be performed. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-05418. The finding is more than minor because it is associated with the protection against external factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors reviewed Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," that states in the Assumptions and Limitations section, The Fire Protection SDP focuses on risks due to degraded conditions of the fire protection program during full power operation of a nuclear power plant. This tool does not address the potential risk significance of fire protection inspection findings in the context of other modes of plant operation (i.e., low power or shutdown). Therefore, the senior reactor analyst evaluated the finding in accordance with Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for both PWRs and BWRs. The finding did not require a quantitative assessment because adequate mitigating equipment remained available; the finding did not increase the likelihood of a loss of reactor coolant system inventory; the finding did not degrade the ability to terminate a leak path or add reactor coolant system inventory; and the finding did not degrade the ability to recover decay heat removal if lost. Therefore, the finding screened as Green, having very low safety significance. The inspectors determined that the apparent cause of this finding was that site management did not ensure that hot-work supervisors were engaged in ensuring compliance with procedural requirements. This finding had a cross-cutting aspect in the area of human performance associated with work practices component because the licensee failed to ensure supervisory oversight of hot-work activities is performed within procedural requirements such that nuclear safety is supported [H.4(c)] (Section 40A3) Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: Self-Revealing Item Type: FIN Finding Page 1 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Manual Reactor Scram Caused by Failure to Ensure the Main Steam Supply Valve to Reactor Feed Pump Turbine B was Full Open Green . The inspectors reviewed a Green self-revealing finding for the failure to ensure the correct position (full open) of the main steam supply valve 1N11-F014B to reactor feed pump turbine B, which resulted in a manual reactor scram due to decreasing reactor water level. During plant shutdown activities to begin refueling outage 18, the at-the-controls operator manually scrammed the reactor from approximately 23 percent rated thermal power due to the decreasing reactor water level. Water level in the reactor was decreasing because valve 1N11-F014B was not fully open, and because pressure in the main steam lines had been reduced when the crew opened turbine bypass valves to begin cooling the main turbine. With valve 1N11-F014B less than fully open and reduced steam pressure, the operating feed pump wasnt able to maintain water level. After the scram, reactor core isolation cooling and reactor feed pump turbine A were used to restore water level. The licensee plans to repair valve 1N11-F014B during the current refuelling outage. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01838. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors concluded that the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment would not be available. The inspectors, in consultation with the regional senior reactor analyst, performed a Phase 2 estimation using the pre-solved work sheets for Grand Gulf Nuclear Station. The inspectors determined by entering the power conversion system column that the finding was of very low safety significance (Green). This result was validated by the senior reactor analyst using the current revision of the plant-specific SPAR model. The inspectors determined the finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the operating staff proceeded with the start up of the reactor feed pump B with the main steam supply valve 1N11-F014B in an unknown position [H.1(b)](Section 1R11). Inspection Report# : 2012002 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Online Risk Assessment Per Severe Weather Off Normal Procedure Due to a Declared Tornado Warning Affecting Grand Gulf Nuclear Station Green. The inspectors identified a Green non-cited violation of Technical Specifications 5.4.1.a for the failure to perform an online risk assessment per severe weather off normal procedure due to a declared tornado warning affecting Grand Gulf Nuclear Station. At 7:41 p.m., on February 15, 2012, the National Weather Service issued a tornado warning for Claiborne County, the county in which Grand Gulf Nuclear Station is located. In response to a tornado warning, licensee procedures required them to enter Off-Normal Operating Procedure 05-1-02-VI-2, Severe Weather, and evaluate online risk. This severe weather condition would have resulted in the licensee entering into an orange risk condition. On February 16, 2012, the inspectors identified that the licensee had not made a log entry for entry into their off normal severe weather procedure during the preceding evening and therefore had not evaluated online risk status for the severe weather condition. In response to the inspectors observations, the licensee initiated a condition report detailing the failures to enter the off normal procedure and enter the correct risk condition. The licensee has implemented short-term corrective actions to ensure the site adequately evaluates the risk associated with adverse weather. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01707. The finding is more-than-minor because it is associated with the Initiating Events Cornerstone attribute of protection against external events, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Page 2 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Appendix K; Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1; Assessment of Risk Deficit; and consulting with the regional senior reactor analyst, the inspectors determined the finding to be of very low safety significance based on a licensees calculated determination of the incremental core damage probability deficit of 4.0E-08. This result was validated by the senior reactor analyst using the current revision of the plant-specific SPAR model. The inspectors determined the finding has a cross-cutting aspect in the area of human performance associated with the resources component because the on-shift senior reactor operators did not have adequate access to current weather information that would prompt control room personnel to re-evaluate risk due to changing weather conditions [H.2(d)](Section 1R13). Inspection Report# : 2012002 (pdf) Mitigating Systems Significance: Sep 21, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Operability Determinations Green. The inspectors identified two examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure EN-OP-104, Operability Determinations. Specifically, for Condition Report CR-GGN-2012-09690, which documents an oil leak on the standby liquid control pump B, and for Condition Report CR-GGN-2012-09889, which documents degraded bolts on a flanged connection on standby service water B piping, the licensee failed to validate that operability evaluations completed for prior non-conforming conditions bounded the conditions documented in the new condition reports. As immediate corrective actions, the licensee re-performed the evaluations and established an adequate basis for operability for the conditions described in the two condition reports listed above. The licensee entered this issue into their corrective action program as CR-GGN-2012-09735 and CR-GGN-2012-10664. The finding was more than minor because if left uncorrected, not performing operability determinations in accordance with procedure could lead to a more significant safety concern. Specifically, if a condition renders a safety related system inoperable and because of this performance deficiency the licensee incorrectly determines that the system is operable, then this performance deficiency could result in a safety related system remaining inoperable for a long period of time. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because although it affected the design or qualification of a mitigating system, the system maintained its operability. The finding had a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly evaluate for operability conditions adverse to quality [P.1(c)] (Section 1R15). Inspection Report# : 2012004 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Page 3 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Preconditioning of 4160 Vac Circuit Breakers for As-Found Tests Green. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to July 27, 2012, the licensees preventive maintenance Procedures 07-S-12-41, 07-S-12-42, and 07-S-12-61 failed to assure that the 4160 Vac circuit breakers would perform satisfactorily in service when the licensee performed maintenance prior to completing as-found tests to verify past operability of the circuit breakers. This finding has been entered into licensees corrective action program as Condition Reports CR-GGN-2012-09035 and CR- GGN-2012-9103. The team determined that failure to establish a test program which ensures that test and maintenance procedures associated with safety-related 4160 Vac circuit breakers would perform satisfactorily in service was a performance deficiency. This finding was more than minor because, if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to perform as-found tests prior to performing maintenance in preventive maintenance procedures was a significant programmatic deficiency which could cause unacceptable conditions to go undetected. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee failed to ensure that test and maintenance procedures were complete, accurate, and up-to-date to assure nuclear safety. [H.2(c)] (1R21.2.1) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish a Testing Program for Safety Related 125 Vdc Circuit Breakers Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to July 27, 2012, the licensee failed to establish a test program for 125 Vdc safety related molded case circuit breakers incorporating the requirements of IEEE 308, to ensure the breakers would not degrade and would perform satisfactorily in service. The finding was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-09030 and CR-GGN-2012-09175. The team determined that the failure to establish a testing program incorporating the requirements of IEEE 308 was a performance deficiency. The finding was more than minor, because if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to establish a testing program was a significant programmatic deficiency that would lead to missed opportunities to detect potential common cause failures from degradation of performance in more than one redundant safety division. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component; because the licensee failed to thoroughly evaluate problems such that resolutions address cause and extent of condition. Specifically, the licensee failed to thoroughly evaluate the extent of condition associated with previously identified NRC violation involving the failure to test 480 Vac molded case circuit breakers identified during the 2009 component design basis inspection. [P.1(c)] (1R21.2.2) Page 4 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for a Change to Credible Passive Failures in the Standby Service Water System Severity Level IV. The team identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests and Experiments which states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if this activity would; result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report (as updated). Specifically, on August 18, 1987, the licensee implemented a change to the updated safety analysis report which limited credible passive failures in the standby service water system to pump and valve seal leakage without obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR GGN 2012 09267. The team determined that the licensees failure to receive prior NRC approval for changes in licensed activities regarding single passive failure criteria for the standby service water system was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding had the ability to impact the regulatory process. The performance deficiency was more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation. In accordance with the NRC Enforcement Manual, risk insights from the Inspection Manual Chapter 0609, Significance Determination Process, are used in determining the significance of 10 CFR 50.59 violations. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because, although the standby service water system could not provide 30 days of decay heat removal without operator action to provide makeup water to the system, it would have been able to complete its 24-hour risk significant mission time. Since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. The finding does not have a crosscutting aspect because the most significant contributor to the finding does not reflect current licensee performance. (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Enter an NRC Violation Regarding the Standby Service Water System into the Corrective Action Program Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as Page 5 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, on July 12, 2012, the NRC informed the licensee of a violation of 10 CFR 50.59 requirements, but the licensee failed to promptly identify this as an adverse condition and enter this condition into their corrective action program until July 19, 2012. The finding was entered into the licensees corrective action program as CR-GGN-2012-10075. The team determined that the licensees failure to promptly enter the NRC violation as condition adverse to quality into the corrective action program was a performance deficiency. This finding was more than minor because it adversely affected the design control attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to promptly document a violation of 10 CFR 50.59, which delayed an operability evaluation that ultimately determined that compensatory measures were required to ensure that the standby service water system could perform its specified safety function for its entire mission time. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because, although the standby service water system could not provide 30 days of decay heat removal without operator action to provide makeup water to the system, it would have been able to complete its 24-hour risk significant mission time. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated, and that actions are taken to address safety issues, in a timely manner, commensurate with their safety significance. Specifically, the licensee did not implement a corrective action program with a low threshold for identifying issues completely, accurately, and in a timely manner commensurate with their safety significance. [P.1(a)] (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Operability Determination Process Procedure Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings which states, in part, that Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, from July 19, 2012, to July 29, 2012, the licensee failed correctly evaluate the operability of the standby service water system with a degraded or nonconforming condition and failed to document a sound basis for a reasonable expectation of operability of the standby service water system as required by Procedure EN-OP-104, Operability Determination Process. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2012-09356. The team determined that the failure to implement the requirements of the operability determination process procedure was a performance deficiency. The finding was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of Page 6 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 systems that respond to initiating events to prevent undesirable consequences. Specifically, the standby service water system was incapable of performing its specified safety function for the entire 30-day mission time without compensatory measures. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because the standby service water system could would have been able to complete its 24-hour risk significant mission time although it could not provide 30 days of decay heat removal without operator action to provide makeup water to the system. This finding had a crosscutting aspect in the area of human performance, decision making component, because the licensee did not make decisions that demonstrated that nuclear safety was an overriding priority. Specifically, the licensee did not make safety significant decisions using a systematic process to ensure safety is maintained. [H.1(a)] (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Test and Inspection Requirements for 4160 Vac Circuit Breakers into Preventive Maintenance Procedures Green. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design document. Specifically, prior to July 27, 2012, the licensees safety-related 4160 Vac circuit breaker preventive maintenance Procedures 07-S-12-41, 07-S-12-42, and 07 S 12-61 failed to incorporate inspection and test requirements for minimum voltage tests, reduced voltage tests, and inspection of auxiliary switch relay contacts as established in the licensees circuit breaker maintenance program. This condition was entered into the licensees corrective action program as Condition Reports CR GGN 2012-08885 and CR-GGN-2012-09111. The team determined that the failure to incorporate required tests and inspections into preventive maintenance procedures for safety related 4160 Vac circuit breakers was a performance deficiency. This finding was more than minor because, if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to incorporate the testing, cleaning, and inspection requirements into preventive maintenance procedures were a significant programmatic deficiency which could cause unacceptable conditions to go undetected. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of problem identification and resolution, operating experience component, because the licensee failed to use operating experience information, including vendor recommendations and internally generated lessons learned, to support plant safety. Specifically, the licensee did not implement and institutionalize operating experience through changes to processes, procedures, equipment, and training programs. [P.2(b)] (1R21.2.4) Page 7 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Inspection Report# : 2012008 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: FIN Finding Failure to Ensure Materials are Stored Properly in the 500 KV Switchyard The inspectors identified a finding for the licensees failure to ensure that materials or equipment were not stored under energized lines or near energized equipment in accordance with station procedures. On May 21, 2012, the inspectors were performing a grid stability inspection and toured the 500 KV switchyard with the system switchyard engineer. During the tour, the inspectors identified numerous cylindrical shaped items stored under a 500 KV power line, which posed a missile hazard to the offsite source of power. The licensee determined that the items in question were bushing sleeves that were left in the switchyard following 500 KV breaker maintenance. The inspectors researched station procedures and determined that the cylindrical items stored under the energized 500 KV power line did not meet procedure requirements for the storage of materials and equipment. Immediate corrective actions included having the items removed from the switchyard. The licensee entered this issue into their corrective action program as Condition Report CR-GGNS-2012-07362. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigation Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors reviewed Manual Chapter 0609, Attachment A, Phase 1 - Initial Screening and Characterization of Findings. Attachment A, Table 4.a, states that a Phase 3 is required if the finding is potentially risk significant due to external initiating event core damage accident sequences. The inspectors determined that the failure to properly store the bushing sleeves in the switchyard could have resulted in a loss of offsite power during a severe weather initiating event. Therefore, the senior reactor analyst evaluated the finding to determine its significance using hand calculations and the site-specific SPAR model. The analyst determined that the probability of having straight-line winds or winds generated by hurricanes or tornados that were strong enough to throw the bushing sleeves into switchyard electrical equipment was between 2.5 x 10-1 and 2.0 x 10-2 /year. The analyst also determined that the conditional probability that bushing sleeves thrown by winds would result in a loss of offsite power was between 1.2 x 10-1 and 1.1 x 10-7. Finally, the SPAR model calculated that the conditional core damage probability for a loss of offsite power initiated in the switchyard was 5.3 x 10-5. Using these values, under all scenarios evaluated by the analyst, the change in core damage frequency caused by the subject performance deficiency was below 1 x 10-6. Therefore, the finding was of very low safety significance (Green). The inspectors determined the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement the corrective action program with a low threshold for identifying materials improperly stored in the 500 KV switchyard [P.1(a)](Section 1R01). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Loss of Alternate Method of Decay Heat Removal Due to Reactor Water Clean Up Pumps Tripping on Low Suction Flow Signal Green. The inspectors reviewed a self-revealing finding for the licensees failure to identify that de-energizing non-safety electrical bus 13BD1 and 13BD2 would cause the reactor water clean-up pumps A and B to trip on a low suction flow signal. On April 24, 2012, the plant was shut down for refueling outage 18, the residual heat removal system B was in service, and the reactor water clean-up system was in standby mode as the alternate shutdown cooling system. In this configuration, the plant was in yellow risk due to having two available systems for decay heat removal. At 10:00 a.m., both reactor water clean-up pumps tripped on low pump suction flow, causing the plant to enter an unplanned orange risk configuration for only having one system available for decay heat removal. The Page 8 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 licensee determined the reactor water pumps tripped while opening the feeder breaker for the 13BD1 and 13BD2 buses (breaker 152-1305) for scheduled maintenance. When breaker 152-1305 was opened, optical isolator AT12 caused the pump low suction flow trip control contacts to close, which initiated the low suction flow alarm and caused the pumps to trip. Immediate corrective actions included restoring reactor water clean-up as the alternative source of decay heat removal by closing breaker 152-1305 and re-energizing the 13BD1 and 13BD2 buses. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2012-06092 and CR-GGN-2012-06105. The finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and it affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a loss of a system safety function. The inspectors determined that the cause of this finding was a latent issue; therefore no cross-cutting aspect was assigned (Section 1R13). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions to Address Configuration Control of Previous Non-cited Violation Green. The inspectors identified a non-cited violation of very low safety significance of 10 CFR Appendix B, Criterion XVI, "Corrective Action," for failure to implement adequate corrective actions for a previous NRC-identified non-cited violation. The previous finding involved a failure to maintain configuration control of various systems in the plant. In response to the previous finding, the licensee performed an apparent cause evaluation and developed actions to address the causes and extent of condition. However, the inspector identified that the actions pertaining to the extent of condition were not properly implemented and, as a result, the deficiency identified by the inspector was not fully resolved. The licensee failed to identify brass compression fittings installed on drain tailpieces of the standby service water system instead of stainless steel fittings as required by design documents. Furthermore, the licensee failed to update applicable design drawings allowing sacrificial compression fittings to be installed. The licensee performed corrective actions to restore configuration control. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-04003, CR-GGN-2012-4180, and CR-GGN-2012-04233. The issue is more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, the issues identified by the inspector impacted the licensees ability to establish and maintain configuration control for equipment relied on for safe operation of the plant. The design control attribute of the Mitigating Systems Cornerstone and the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences were affected. Until the issues are fully resolved, the licensee continues to be vulnerable to gaps in their system configuration control. The finding was determined to be of very low safety significance (Green) using Attachment 4 to IMC 0609, "Significance Determination Process," because it did not result in an actual loss of safety function. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance associated with the resources component because the licensee did not provide adequate training of personnel so that the inappropriately installed fittings could be identified during system walkdowns [H.2(b)] (Section 1R08). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement a Surveillance Requirement to Assure that the Limiting Condition for Operation Will be Page 9 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Met Green. The inspectors identified a non-cited violation of 10 CFR Part 50.36, Technical Specifications, involving the failure to implement a surveillance requirement to assure that the limiting conditions for operation of the ultimate heat sink will be met. Technical Specifications requires two cooling towers and two cooling basins, with the volume of the two basins constituting the entire inventory of the ultimate heat sink. Therefore, an interconnecting siphon line is installed to transfer water between the two cooling tower basins. That siphon line has the safety-related function of ensuring the availability of enough cooling water to satisfy ultimate heat sink requirements. Technical Specification 3.7.1 includes Surveillance Requirement 3.7.1.1, which verifies the water level in each cooling tower basin every 24 hours, and Surveillance Requirement 3.7.1.2, which verifies each cooling tower fan every 31 days. However, the inspectors identified that Technical Specification 3.7.1 does not include a surveillance requirement to verify that the interconnecting siphon line will perform its safety-related function. On May 20, 2012, the licensee performed an operability test for the siphon line and determined that it was operable. The licensee is currently performing a preventative maintenance task as a compensatory action to ensure operability of the siphon line until a license amendment can be submitted to the NRC that establishes a surveillance requirement. The licensee documented this violation in Condition Reports CR-GGN-2012-08257 and CR-GGN-2012-08537. The violation is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, without a surveillance requirement that verifies the interconnecting siphon line can perform its safety-related function, the licensee cannot ensure that sufficient cooling water is available following an accident. The inspectors evaluated the finding using Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings and determined that the finding was of very low safety significance (Green) because the finding was a design or qualification deficiency confirmed not to result in a loss of operability or function; did not represent a loss of safety system function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time; 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 in the human performance area associated with the resources component because the licensee did not ensure that equipment was adequate to assure nuclear safety, in that the licensee had recently reviewed documentation associated with a modification to the siphon line but failed to identify that operability of the UHS could not be established without a technical specification surveillance requirement to ensure operability of the siphon line [H.2(c)] (Section 1R19). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow a Post-Modification Test Procedure Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to follow a post-modification test procedure for the interconnecting siphon line between the two standby service water system cooling tower basins. Operability of the ultimate heat sink is based on a minimum water level in the two standby service water cooling tower basins, an operable interconnecting siphon between the basins, and four operable cooling tower fans (two per basin). At extended power uprate conditions, the configuration of the basins and the original siphon line would not support 30 days of operation of both trains of the standby service water system and the high pressure core spray service water systems without makeup, so the licensee performed a modification (EC 25649), which involved replacing the original siphon line with a new siphon line in order to transfer water from one basin to the other. On March 28, 2012, after completing the modification, the licensee performed post-modification testing to determine the piping friction loss coefficient of the modified siphon line and to evaluate its acceptability against the worst-case friction loss coefficient documented in EC 25649. The licensee deviated from the test procedure, as-written, and performed the test with an inadequate pressure gauge instead of the specified gauge. After inspectors challenged the validity of these test results, the licensee performed another test of the siphon line with a different Page 10 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 method that did not require the use of a pressure gauge to measure the piping friction loss coefficient. The inspectors reviewed the subsequent test data and found the test results to be satisfactory. The licensee documented this concern in Condition Report CR-GGN-2012-05260. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the use of an unqualified gauge invalidated the test results, and a different test method had to be developed to determine the piping friction loss coefficient for the siphon line. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed to result in loss of operability or function; did not represent a loss of safety system function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time; 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 in the human performance area associated with work practices component because licensee personnel proceeded in the face of uncertainty or unexpected circumstances. Specifically, the licensee proceeded with the test without verifying that the pressure gauge was suitable for the test conditions after observing unexpected measurements with the gauge [H.1(a)] (Section 1R19). Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions Associated with Division 1 and 2 Standby Service Water Safety Related Cables that were Partially Submerged in Cable Manhole/Vault Green . The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take timely corrective actions to correct a condition adverse to quality associated with division 1 and 2 standby service water safety related cables that were partially submerged in a cable manhole/vault. The inspectors reviewed work order 52284535 and noted that the sump pump for manhole/vault MH-01, which contained standby service water cables for division 1 and 2, was determined to be non-functional on September 10, 2011. The inspectors determined that a work order to repair the non-functioning sump pump had been developed but that the work order had not yet been scheduled. During a subsequent inspection, manhole/vault MH-01 was found to contain approximately three feet of water, with water partially covering some of the safety related cables. The electricians immediately pumped manhole/vault MH-01 and wrote a condition report. The licensee repaired the sump pump the next week and declared it functional. The cables remained operable based on the results of meggar tests. The licensee entered this issue into their corrective action program as condition reports CR-GGN-2012-00503, 01324, and 01389. The finding is more than minor because it is associated with the equipment performance attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent either a loss of system safety function or an actual loss of safety function of a single train of one or more non-Technical Specification 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 in the area of human performance associated with the work practices component, in that the licensee personnel did not initiate a condition report as required by licensee procedure when the work order associated with sump pump testing of MH-01 determined that the sump pump was not functioning properly [H.4(b)] (Section 1R06). Inspection Report# : 2012002 (pdf) Page 11 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Condition Adverse to Fire Protection, in That the Licensee Failed to Adequately Provide Contingency Lighting in the Fire Brigade Dress Out Area While Normal Lighting was Inoperabl Green. The inspectors identified a Green non-cited violation of Facility Operating License Condition 2.C(41), for the failure to correct a condition adverse to fire protection. Specifically, the licensee failed to adequately provide contingency lighting in the fire brigade dress out area while normal lighting was inoperable due to maintenance on an associated breaker. The inadequate lighting delayed fire brigade response to a potential fire in the turbine building. Immediate corrective action included placing temporary lighting in the area. Normal lighting to the area was restored the next week. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01488. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined from table 3b that issues related to performance of the fire brigade are not included in Appendix F and require NRC management review using Appendix M. Regional management review evaluated the overall impact of lighting issue in the fire brigade dress out area and concluded that, while the fire protection defense-in-depth was affected by the performance deficiency, the overall defense-in-depth of the front-line systems was not impacted because of train separation and safe shutdown analysis at the site. Therefore the finding screened as having very low safety significance (Green) in accordance with Manual Chapter 0609, Appendix M. The inspectors determined the finding had a cross-cutting aspect in the area of human performance associated with the work control component, in that licensee personnel failed to ensure adequate job site conditions (lighting in the fire bridge dress out area) were in place prior to performance electrical maintenance in the turbine building [H.3(a)] (Section 40A3). Inspection Report# : 2012002 (pdf) Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Inspection of Probable Maximum Precipitation Door Seals Protecting Safety Related Equipment Green. The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to perform an adequate inspection of probable maximum precipitation door seals protecting safety related equipment. Inspectors found that one of the door seals to standby service water pump house A was in a degraded condition. The inspectors identified that the door seal did not make complete contact with the door frame all the way around. The licensee determined that the probable maximum precipitation seal for the identified door was in a degraded condition. Failure of this door seal during a probable maximum precipitation event could potentially cause flooding of the standby service water pump house A. Immediate corrective actions included the site initiating compensatory actions for the degraded seal by staging sand bags in the area and requiring monitoring of the affected door during heavy rainfall. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2011-07687. The finding is more than minor because it is associated with the protection against external factors attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors used the seismic, flooding, and severe weather Table 4b and determined that it would not affect multiple trains of safety equipment and that the finding had very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the resources component in that the licensees procedure used for the inspection of the Page 12 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 door seals did not take into account the status of the pump house ventilation system while performing the door seal inspection, and therefore, the licensee failed to make the required adjustments to the door seals resulting in their inspections of the probable maximum precipitation door seals being inadequate [H.2(c)] (Section 1R05). Inspection Report# : 2011005 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions for Reactor Core Isolation Cooling System Venting Green. The team identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct an inadequate venting procedure for the reactor core isolation cooling system. Corrective actions were not taken in a timely enough manner such that resolution was reached prior to time to demonstrate the licensee met their applicable technical specification surveillance requirement. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-07669 and subsequently altered their procedure, which performs the technical specification surveillance requirement to demonstrate that it meets the applicable requirements. This finding is more than minor because it affects the procedure quality 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. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the decision making component. The licensee did not use conservative assumptions when deciding to pursue corrective action for venting of the reactor core isolation cooling system piping to demonstrate their action was safe in order to proceed rather than demonstrating it was unsafe to disapprove the action [H.1(b)]. (Section 4OA2.5a) Inspection Report# : 2011006 (pdf) Significance: SL-IV Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Submit a Licensee Event Report for an Inoperable Reactor Core Isolation Cooling System Severity Level IV. The team identified a Severity Level IV noncited violation of 10 CFR 50.73, Licensee Event Report System, associated with the licensees failure to submit a licensee event report within 60 days following discovery of an event meeting the reporting criteria as specified. Specifically, the licensee was not meeting the technical specification surveillance requirement for venting the reactor core cooling isolation system and subsequently the system was inoperable in excess of the allowed outage time which constituted a condition prohibited by technical specifications. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8890. This finding affects the mitigating systems cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRCs ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section 6.9 of the Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. This finding has no crosscutting aspect, as it is not indicative of current performance (Section 4OA2.5b). Inspection Report# : 2011006 (pdf) Page 13 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Significance: Oct 21, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Document a Condition as a Significant Condition Adverse to Quality Green. The team identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to identify and document a significant condition adverse to quality and report the condition to appropriate levels of management. As a result, a root cause analysis was not performed and more comprehensive actions to prevent recurrence were not considered for the condition. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011- 07671, to address the problem. This finding is more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance (Green) because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensees procedures for significant conditions adverse to quality were not complete and accurate enough to prevent the condition. [H.2(c)]. (Section 4OA2.5c) Inspection Report# : 2011006 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: VIO Violation Inadequate Corrective Action for a Leak on the Division II Emergency Diesel Generator Lube Oil Sump Green. The team identified a Green cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a leak on the Division II emergency diesel generator lube oil sump. Despite identification of the leak in 2004, ineffective attempts to repair the leak and previous identification by the NRC in 2009, the licensee dispositioned the leak as accept as-is without a full understanding of the lube oil sump leak and potential consequences. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8880. The condition was discovered and documented by the licensee in 2004. This finding was initially determined by the NRC to be a minor violation in 2009. Paragraph F of Section 2.10 of the NRC Enforcement Manual states in part that where a licensee does not take corrective action for a minor violation, the matter should be considered more than minor and associated with a green inspection finding and dispositioned in a cited or noncited violation, as appropriate. This finding is now determined to be more than minor because if left uncorrected the failure to restore the lube oil sump for the Division II emergency diesel generator to design conditions would have the potential to lead to a more significant safety concern, specifically, the leak could worsen and potentially affect operability of the emergency diesel generator. Due to the licensees failure to restore compliance within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy. This finding affects the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate this problem such that the resolutions addressed the causes [P.1(c)]. (Section 4OA2.5d) Page 14 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Inspection Report# : 2011006 (pdf) Barrier Integrity Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Loss of Decay Heat Removal to the Spent Fuel Pool Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specifications 5.4.1(a), involving a loss of decay heat removal in the spent fuel pool due to station personnel failing to correctly follow operation of pool gate seal air supply procedure. On April 17, 2012, Grand Gulf Nuclear Station was preparing to drain the reactor cavity to reinstall the vessel head after the completion of refueling activities. In preparation, the upper containment pool to the reactor cavity gate was installed by General Electric-Hitachi technicians with Entergy oversight. Technicians were directed by procedure to verify that all supply isolation toggle valves to the gate seals were open and secured in place. However, technicians failed to complete this action correctly and the control room was informed that all prerequisites were completed and began the cavity drain down. The control room immediately noticed the fuel pool drain tank level was decreasing and attempted to makeup to the tank via the normal makeup valve. When the fuel pool drain tank level reached 17 percent full, both fuel pool cooling and cleanup pumps tripped as expected, resulting in loss of decay heat removal to the spent fuel pool. The main control room entered the off-normal event procedure for inadequate decay heat removal, and they secured the drain down evolution. Approximately 47 minutes later, spent fuel pool cooling was re-established. During this event, the spent fuel pool temperature did not exceed the limits required by Technical Requirements Manual Section 6.7.4 (140°F). Short term corrective actions included restoring decay heat removal to the spent fuel pool and conducting a human performance review of the event. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-05756. The finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represented a loss of spent fuel pool cooling that would not preclude restoration of cooling to the spent fuel pool prior to pool boiling. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to use adequate self- and peer-checking techniques to ensure gate seals were properly inflated prior to cavity drain down [H.4(a)] (Section 1R20). Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Modification of the Spent Fuel Pool without Prior NRC Approval SLIV. The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests and Experiments, when the licensee failed to obtain a license amendment prior to implementing a proposed change to the plant that required a change to Technical Specifications. The 10 CFR 50.59 evaluation performed by the licensee is dated January 24, 2001, thus it was performed under the requirements of the old rule based on the Entergy Operations letter dated March 5, 2001. In the 10 CFR 50.59 evaluation for the removal of Blackness Testing and the division of the spent fuel pool into two regions, the licensee determined that the modifications did not require a change to Page 15 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Technical Specifications. However, 10 CFR 50.36, Technical Specifications, Section 4, Design Features, requires that design features such as geometric arrangements, which, if altered or modified, would have a significant effect on safety, must be incorporated into Technical Specifications. The NRC considers that the establishment of two regional zones in the spent fuel pool, each having specific loading criteria to maintain keff less than 0.95, constitutes design features which, if altered or modified would have a significant effect on safety. Therefore, these design features should have been incorporated into the Technical Specifications. In a letter dated September 8, 2010, (ML102660403), the licensee submitted a power up-rate license amendment request. The NRC staff is currently reviewing the license request, which includes the licensees technical justification for the spent fuel pool changes described above. Based on preliminary review of the amendment request, the NRC staff has determined that an immediate safety concern does not exist. The licensee has entered this issue into their corrective action program as condition report CR-GGN-2012-01077. The finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, containment) protect the public from radionuclide releases caused by accidents or events. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characteristics of Findings, and determined that the finding was of very low safety significance (Green) because it did not result in the loss of cooling to the spent fuel pool, did not result from fuel handling errors that caused damage to fuel clad integrity, and it did not result in a loss of spent fuel pool inventory. This finding is a latent issue and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 1R15). Inspection Report# : 2012002 (pdf) Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary. Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Page 16 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Use of Waivers to Allow Workers to Exceed the Minimum Day Off Rule Green. The inspectors identified a non-cited violation of 10 CFR 26, Subpart I, Managing Fatigue, Subsection 207, Waivers and Exceptions, when the licensee inappropriately used waivers to allow workers to exceed the minimum day off rule. While reviewing condition reports, the inspectors noted the use of work hour waivers for a large number of staff. The circumstances for the use of waivers were the refueling outage lasting more than 60 days, contract expiration leading to 14 layoffs, and the loss of 4 workers via voluntary resignation. Due to these circumstances, work hours and fatigue of waivered individuals would have to be assessed daily. The assessment is required because the work hour limit of these individuals exceeded the minimum day off rule, therefore requiring daily monitoring until the end of the cycle. The waivered individuals averaged two days off per six-week period compared to the required three days off. Title 10 CFR 26.207 (a)(2) allows the granting of waivers only to address circumstances that could not have been reasonably controlled. The inspectors determined that the licensee was aware of the circumstances of an extended refueling outage and contract renewal deadline well in advance of the need to grant waivers, and a reasonable amount of time was available for the licensee to develop and execute contingency plans to negate the need to use waivers. Corrective actions included initiating assessments and waivers for exceeding minimum days off requirements for shift personnel for the six-week period ending May 27, 2012, and returning to the normal on-line work schedule in which adequate manpower is available to meet the requirements of the rule. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-7348. The finding is more than minor because it is associated with the access authorization attribute of the Security Cornerstone, and affected the cornerstone objective to provide assurance that the licensees security system and material control and accounting program use a defense in-depth approach and can protect against (1) the design basis threat of radiological sabotage from external and internal threats, and (2) the theft or loss of radiological materials. Using the Inspection Manual Chapter 0609, Appendix E, Baseline Security Significance Determination Process for Power Reactors, Figures 5 and 6, the finding was determined to have very low security significance because the calculated point total did not exceed the threshold value for a Green non-cited violation. The cumulative total for this finding was zero points, which was calculated by factoring the one impact area (vital areas) against Tier III Element 08.02.08, security force work hours, of the access authorization attribute, which resulted in a total of zero points within this attribute. The finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision making component in that the licensee failed to use conservative assumptions in developing staff schedules for the duration of refueling outage 18 and for allowing an employment contract to expire that led to 14 individuals being laid off without realizing the impact these decisions would have on the licensees ability to meet the requirements of the rule [H.1(b)] (Section 1R20). Inspection Report# : 2012003 (pdf) Significance: N/A Dec 01, 2011 Identified By: NRC Item Type: FIN Finding Grand Gulf, 2011, Biennial Problem Identification and Resolution Inspection Summary The inspectors concluded that the licensee was, in general, effective in identifying, evaluating, and resolving problems. Grand Gulf personnel were identifying and entering issues into the corrective action program at appropriately low thresholds as evidenced by a large number of condition reports issued. The team determined that the licensee generally screened issues appropriately for operability and reportability. The team noted that issues were typically identified promptly and prioritized commensurate with their safety significance. Most root and apparent cause analyses appropriately considered extent of condition and previous occurrences. The team concluded that the corrective actions were generally identified and implemented promptly. The team found that the licensee had established and was maintaining an environment at Grand Gulf where employees felt free to raise safety concerns without fear of retaliation. Page 17 of 18

3Q/2012 Inspection Findings - Grand Gulf 1 The licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items in the corrective action program. The licensee used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective quality assurance audits and self assessments, as demonstrated by self identification of corrective action program areas for improvement. Inspection Report# : 2011006 (pdf) Last modified : November 30, 2012 Page 18 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 Grand Gulf 1 4Q/2012 Plant Inspection Findings Initiating Events Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Risk Significances and Develop Action Plans to Address Equipment Identified During Extent of Condition Review for a Post Scram Root Cause Analysis Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to follow procedure EN-LI-118, Root Cause Evaluation Process, Revision 18, in that they failed to evaluate the risk significances and develop action plans to address equipment identified during their extent-of-condition review for a post-scram root cause analysis. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11950. The immediate corrective actions included assigning corrective actions for operations personnel to properly evaluate the risk significance of the identified components and perform appropriate corrective actions to correct the degraded conditions. The licensees failure to properly determine risk significance and associated action plans to correct degraded equipment that could challenge safe plant operation is a performance deficiency. The performance deficiency is more than minor and is therefore a finding because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to take corrective actions to correct degraded equipment has the potential to lead to initiating events resulting in plant transients. Using NRC Inspection Manual Chapter 0609, , "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the apparent cause of this finding was that when operations management directed operators to identify the degraded equipment, they did not encourage those operators to comply with Procedure EN-LI-118. Therefore, the finding has a cross-cutting aspect in the human performance area, work practices component because the licensee did not define and effectively communicate expectations regarding procedural compliance. [H.4 (b)] (Section 4OA3). Inspection Report# : 2012005 (pdf) Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of Hot Work Fire Watch to Follow Procedural Requirements The inspectors reviewed a self-revealing non-cited violation of Technical Specifications 5.4.1(a), for failure of the hot-work fire watch to follow procedural requirements, which resulted in a fire in main condenser A. On April 11, 2012, at 6:11 p.m., hot-work was in progress inside the condenser A in the upper southeast corner at 150 foot elevation. Cutting was being performed by contract boilermakers using an oxy-acetylene torch, with ventilation exhaust and supply provided by nearby HEPA hoses. The torch cutting operation produced hot slag, which exited the barrier provided by the fire blankets and ignited the nearby HEPA hoses, air conditioning hoses, and eventually the acetylene hoses. Contract pipefitters in the area were able to extinguish the fire. The main control room was informed of the fire inside condenser A and dispatched the fire brigade to the scene. The operations shift manager declared a notice of unusual event at 6:26 p.m. due to a fire in the protected area lasting longer than 15 minutes. Members of the fire brigade entered the condenser bay at 6:42 p.m. and reported to the control room there was no fire present, only smoke. The notice of unusual event was exited at 7:00 p.m. Short term corrective actions included site management Page 1 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 placing a stop work order on all hot-work until a complete investigation of the event could be performed. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-05418. The finding is more than minor because it is associated with the protection against external factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors reviewed Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," that states in the Assumptions and Limitations section, The Fire Protection SDP focuses on risks due to degraded conditions of the fire protection program during full power operation of a nuclear power plant. This tool does not address the potential risk significance of fire protection inspection findings in the context of other modes of plant operation (i.e., low power or shutdown). Therefore, the senior reactor analyst evaluated the finding in accordance with Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for both PWRs and BWRs. The finding did not require a quantitative assessment because adequate mitigating equipment remained available; the finding did not increase the likelihood of a loss of reactor coolant system inventory; the finding did not degrade the ability to terminate a leak path or add reactor coolant system inventory; and the finding did not degrade the ability to recover decay heat removal if lost. Therefore, the finding screened as Green, having very low safety significance. The inspectors determined that the apparent cause of this finding was that site management did not ensure that hot-work supervisors were engaged in ensuring compliance with procedural requirements. This finding had a cross-cutting aspect in the area of human performance associated with work practices component because the licensee failed to ensure supervisory oversight of hot-work activities is performed within procedural requirements such that nuclear safety is supported [H.4(c)] (Section 40A3) Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: Self-Revealing Item Type: FIN Finding Manual Reactor Scram Caused by Failure to Ensure the Main Steam Supply Valve to Reactor Feed Pump Turbine B was Full Open Green . The inspectors reviewed a Green self-revealing finding for the failure to ensure the correct position (full open) of the main steam supply valve 1N11-F014B to reactor feed pump turbine B, which resulted in a manual reactor scram due to decreasing reactor water level. During plant shutdown activities to begin refueling outage 18, the at-the-controls operator manually scrammed the reactor from approximately 23 percent rated thermal power due to the decreasing reactor water level. Water level in the reactor was decreasing because valve 1N11-F014B was not fully open, and because pressure in the main steam lines had been reduced when the crew opened turbine bypass valves to begin cooling the main turbine. With valve 1N11-F014B less than fully open and reduced steam pressure, the operating feed pump wasnt able to maintain water level. After the scram, reactor core isolation cooling and reactor feed pump turbine A were used to restore water level. The licensee plans to repair valve 1N11-F014B during the current refuelling outage. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01838. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors concluded that the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment would not be available. The inspectors, in consultation with the regional senior reactor analyst, performed a Phase 2 estimation using the pre-solved work sheets for Grand Gulf Nuclear Station. The inspectors determined by entering the power conversion system column that the finding was of very low safety significance (Green). This result was validated by the senior reactor analyst using the current revision of the plant-specific SPAR model. The inspectors determined the finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the operating staff proceeded with the start up of the reactor feed pump B with the main steam supply valve 1N11-F014B in an unknown position [H.1(b)](Section 1R11). Inspection Report# : 2012002 (pdf) Significance: Mar 23, 2012 Page 2 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Online Risk Assessment Per Severe Weather Off Normal Procedure Due to a Declared Tornado Warning Affecting Grand Gulf Nuclear Station Green. The inspectors identified a Green non-cited violation of Technical Specifications 5.4.1.a for the failure to perform an online risk assessment per severe weather off normal procedure due to a declared tornado warning affecting Grand Gulf Nuclear Station. At 7:41 p.m., on February 15, 2012, the National Weather Service issued a tornado warning for Claiborne County, the county in which Grand Gulf Nuclear Station is located. In response to a tornado warning, licensee procedures required them to enter Off-Normal Operating Procedure 05-1-02-VI-2, Severe Weather, and evaluate online risk. This severe weather condition would have resulted in the licensee entering into an orange risk condition. On February 16, 2012, the inspectors identified that the licensee had not made a log entry for entry into their off normal severe weather procedure during the preceding evening and therefore had not evaluated online risk status for the severe weather condition. In response to the inspectors observations, the licensee initiated a condition report detailing the failures to enter the off normal procedure and enter the correct risk condition. The licensee has implemented short-term corrective actions to ensure the site adequately evaluates the risk associated with adverse weather. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01707. The finding is more-than-minor because it is associated with the Initiating Events Cornerstone attribute of protection against external events, and it affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Appendix K; Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1; Assessment of Risk Deficit; and consulting with the regional senior reactor analyst, the inspectors determined the finding to be of very low safety significance based on a licensees calculated determination of the incremental core damage probability deficit of 4.0E-08. This result was validated by the senior reactor analyst using the current revision of the plant-specific SPAR model. The inspectors determined the finding has a cross-cutting aspect in the area of human performance associated with the resources component because the on-shift senior reactor operators did not have adequate access to current weather information that would prompt control room personnel to re-evaluate risk due to changing weather conditions [H.2(d)](Section 1R13). Inspection Report# : 2012002 (pdf) Mitigating Systems Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Adequate Procedure Instructions to Perform Preventive Maintenance Requiring The Periodic Replacement of the Control Relays in the GE Magne Blast Circuit Breakers Green. The inspectors reviewed a self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to complete preventive maintenance tasks on the high pressure core spray division III diesel generator output breaker in accordance with the corresponding preventive maintenance task template. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2012-07992. The immediate corrective actions included replacing the failed control relay and restoring operability to the division III diesel generator. The long term corrective actions included revising breaker refurbishment/replacement procedure with directions to replace the control relay and change the procedure frequency to every 10 years versus every 12 years. The inspectors determined that this performance deficiency was more than minor and is therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this failed control relay caused the subject breaker to fail to close during the division III diesel generator monthly surveillance on June 5, 2012. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," to determine that the issue Page 3 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 affected the Mitigating System Cornerstone. Because the finding pertained only to a degraded condition while the plant was shutdown, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Checklist 8, Cold Shutdown or Refueling Operation - Time to Boil > 2 Hours: RCS Level < 23 Above Top of Flange, to determine that the finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add RCS inventory when needed; did not significantly degrade the licensees ability to recover decay heat removal if lost; and did not affect the safety/relief valves (Green). The inspectors determined that the cause of this finding was a latent issue that is not reflective of current performance, therefore no cross-cutting aspect was identified. (Section 1R20.b). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Gain Settings on APRM and LPRM Instruments in Accordance with Design Requirements Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish the gain settings used on the power range neutron monitoring system in accordance with design requirements. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2013-00177. The immediate corrective actions included adjusting gain settings for their average power range monitor (APRM) instruments to indicate actual core thermal power as determined by the heat balance. In additioin, the licensee revised their neutron monitoring procedure to set the initial gains for the average power range monitor to the maximum value to maintain conservative power indication during future startups. They also changed their local power range monitor replacement procedure to use the vendor specified initial gain setting of 3.692 prior to startup. The finding was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect gain settings caused a violation of technical specification 3.0.4 by rendering the APRM Neutron Flux High - Setdown scram function and the Neutron Flux - Upscale, Startup control rod block function inoperable prior to entry into Mode 2. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had very low safety significance (Green) because although the finding affected a single reactor protection system trip signal to initiate a reactor scram, it did not affect the function of other redundant trips or diverse methods of reactor shutdown, did not involve control manipulations that unintentionally added positive reactivity, and did not result in a mismanagement of reactivity by operators. Because the performance deficiency occurred in the past and is not reflective of current licensee performance, this finding was not assigned a cross-cutting aspect. (Section 4OA3). Inspection Report# : 2012005 (pdf) Significance: Sep 21, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Operability Determinations Green. The inspectors identified two examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure EN-OP-104, Operability Determinations. Specifically, for Condition Report CR-GGN-2012-09690, which documents an oil leak on the standby liquid control pump B, and for Condition Report CR-GGN-2012-09889, which documents degraded bolts on a flanged connection on standby service water B piping, the licensee failed to validate that operability evaluations completed for prior non-conforming conditions bounded the conditions documented in the new condition reports. As immediate corrective actions, the licensee re-performed the evaluations and established an adequate basis for operability for the conditions described in the two condition reports listed above. The licensee entered this issue into their corrective action program as CR-GGN-2012-09735 and CR-GGN-2012-10664. Page 4 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 The finding was more than minor because if left uncorrected, not performing operability determinations in accordance with procedure could lead to a more significant safety concern. Specifically, if a condition renders a safety related system inoperable and because of this performance deficiency the licensee incorrectly determines that the system is operable, then this performance deficiency could result in a safety related system remaining inoperable for a long period of time. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because although it affected the design or qualification of a mitigating system, the system maintained its operability. The finding had a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly evaluate for operability conditions adverse to quality [P.1(c)] (Section 1R15). Inspection Report# : 2012004 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of 4160 Vac Circuit Breakers for As-Found Tests Green. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to July 27, 2012, the licensees preventive maintenance Procedures 07-S-12-41, 07-S-12-42, and 07-S-12-61 failed to assure that the 4160 Vac circuit breakers would perform satisfactorily in service when the licensee performed maintenance prior to completing as-found tests to verify past operability of the circuit breakers. This finding has been entered into licensees corrective action program as Condition Reports CR-GGN-2012-09035 and CR- GGN-2012-9103. The team determined that failure to establish a test program which ensures that test and maintenance procedures associated with safety-related 4160 Vac circuit breakers would perform satisfactorily in service was a performance deficiency. This finding was more than minor because, if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to perform as-found tests prior to performing maintenance in preventive maintenance procedures was a significant programmatic deficiency which could cause unacceptable conditions to go undetected. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee failed to ensure that test and maintenance procedures were complete, accurate, and up-to-date to assure nuclear safety. [H.2(c)] (1R21.2.1) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish a Testing Program for Safety Related 125 Vdc Circuit Breakers Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to July 27, 2012, the licensee failed to establish a test program for 125 Vdc safety related molded case circuit breakers incorporating the requirements of IEEE 308, to ensure the breakers would not degrade and would perform satisfactorily in service. The finding was entered into the licensees corrective action Page 5 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 program as Condition Reports CR-GGN-2012-09030 and CR-GGN-2012-09175. The team determined that the failure to establish a testing program incorporating the requirements of IEEE 308 was a performance deficiency. The finding was more than minor, because if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to establish a testing program was a significant programmatic deficiency that would lead to missed opportunities to detect potential common cause failures from degradation of performance in more than one redundant safety division. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component; because the licensee failed to thoroughly evaluate problems such that resolutions address cause and extent of condition. Specifically, the licensee failed to thoroughly evaluate the extent of condition associated with previously identified NRC violation involving the failure to test 480 Vac molded case circuit breakers identified during the 2009 component design basis inspection. [P.1(c)] (1R21.2.2) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for a Change to Credible Passive Failures in the Standby Service Water System Severity Level IV. The team identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests and Experiments which states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if this activity would; result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report (as updated). Specifically, on August 18, 1987, the licensee implemented a change to the updated safety analysis report which limited credible passive failures in the standby service water system to pump and valve seal leakage without obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR GGN 2012 09267. The team determined that the licensees failure to receive prior NRC approval for changes in licensed activities regarding single passive failure criteria for the standby service water system was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding had the ability to impact the regulatory process. The performance deficiency was more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation. In accordance with the NRC Enforcement Manual, risk insights from the Inspection Manual Chapter 0609, Significance Determination Process, are used in determining the significance of 10 CFR 50.59 violations. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because, although the standby service water system could not provide 30 days of decay heat removal without operator action to provide makeup water to the system, it would have been able to complete its 24-hour risk significant mission time. Since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. The finding does not have a crosscutting aspect because the most significant contributor to the finding does not reflect current licensee performance. (1R21.2.3) Page 6 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Enter an NRC Violation Regarding the Standby Service Water System into the Corrective Action Program Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, on July 12, 2012, the NRC informed the licensee of a violation of 10 CFR 50.59 requirements, but the licensee failed to promptly identify this as an adverse condition and enter this condition into their corrective action program until July 19, 2012. The finding was entered into the licensees corrective action program as CR-GGN-2012-10075. The team determined that the licensees failure to promptly enter the NRC violation as condition adverse to quality into the corrective action program was a performance deficiency. This finding was more than minor because it adversely affected the design control attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to promptly document a violation of 10 CFR 50.59, which delayed an operability evaluation that ultimately determined that compensatory measures were required to ensure that the standby service water system could perform its specified safety function for its entire mission time. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because, although the standby service water system could not provide 30 days of decay heat removal without operator action to provide makeup water to the system, it would have been able to complete its 24-hour risk significant mission time. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated, and that actions are taken to address safety issues, in a timely manner, commensurate with their safety significance. Specifically, the licensee did not implement a corrective action program with a low threshold for identifying issues completely, accurately, and in a timely manner commensurate with their safety significance. [P.1(a)] (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Operability Determination Process Procedure Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings which states, in part, that Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, from July 19, 2012, to July 29, 2012, the licensee failed correctly evaluate the operability of the standby service water system with a degraded or nonconforming condition and failed to document a sound basis for a reasonable expectation of operability of the standby service water system as required by Procedure EN-OP-104, Operability Determination Process. The finding Page 7 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 was entered into the licensees corrective action program as Condition Report CR-GGN-2012-09356. The team determined that the failure to implement the requirements of the operability determination process procedure was a performance deficiency. The finding was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the standby service water system was incapable of performing its specified safety function for the entire 30-day mission time without compensatory measures. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because the standby service water system could would have been able to complete its 24-hour risk significant mission time although it could not provide 30 days of decay heat removal without operator action to provide makeup water to the system. This finding had a crosscutting aspect in the area of human performance, decision making component, because the licensee did not make decisions that demonstrated that nuclear safety was an overriding priority. Specifically, the licensee did not make safety significant decisions using a systematic process to ensure safety is maintained. [H.1(a)] (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Test and Inspection Requirements for 4160 Vac Circuit Breakers into Preventive Maintenance Procedures Green. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design document. Specifically, prior to July 27, 2012, the licensees safety-related 4160 Vac circuit breaker preventive maintenance Procedures 07-S-12-41, 07-S-12-42, and 07 S 12-61 failed to incorporate inspection and test requirements for minimum voltage tests, reduced voltage tests, and inspection of auxiliary switch relay contacts as established in the licensees circuit breaker maintenance program. This condition was entered into the licensees corrective action program as Condition Reports CR GGN 2012-08885 and CR-GGN-2012-09111. The team determined that the failure to incorporate required tests and inspections into preventive maintenance procedures for safety related 4160 Vac circuit breakers was a performance deficiency. This finding was more than minor because, if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to incorporate the testing, cleaning, and inspection requirements into preventive maintenance procedures were a significant programmatic deficiency which could cause unacceptable conditions to go undetected. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of problem identification and resolution, operating experience component, because the licensee failed to use operating experience information, including vendor recommendations and internally generated lessons learned, to support plant safety. Specifically, the licensee did not implement and institutionalize operating experience through changes to processes, procedures, equipment, and training programs. [P.2(b)] (1R21.2.4) Page 8 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 Inspection Report# : 2012008 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: FIN Finding Failure to Ensure Materials are Stored Properly in the 500 KV Switchyard The inspectors identified a finding for the licensees failure to ensure that materials or equipment were not stored under energized lines or near energized equipment in accordance with station procedures. On May 21, 2012, the inspectors were performing a grid stability inspection and toured the 500 KV switchyard with the system switchyard engineer. During the tour, the inspectors identified numerous cylindrical shaped items stored under a 500 KV power line, which posed a missile hazard to the offsite source of power. The licensee determined that the items in question were bushing sleeves that were left in the switchyard following 500 KV breaker maintenance. The inspectors researched station procedures and determined that the cylindrical items stored under the energized 500 KV power line did not meet procedure requirements for the storage of materials and equipment. Immediate corrective actions included having the items removed from the switchyard. The licensee entered this issue into their corrective action program as Condition Report CR-GGNS-2012-07362. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigation Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors reviewed Manual Chapter 0609, Attachment A, Phase 1 - Initial Screening and Characterization of Findings. Attachment A, Table 4.a, states that a Phase 3 is required if the finding is potentially risk significant due to external initiating event core damage accident sequences. The inspectors determined that the failure to properly store the bushing sleeves in the switchyard could have resulted in a loss of offsite power during a severe weather initiating event. Therefore, the senior reactor analyst evaluated the finding to determine its significance using hand calculations and the site-specific SPAR model. The analyst determined that the probability of having straight-line winds or winds generated by hurricanes or tornados that were strong enough to throw the bushing sleeves into switchyard electrical equipment was between 2.5 x 10-1 and 2.0 x 10-2 /year. The analyst also determined that the conditional probability that bushing sleeves thrown by winds would result in a loss of offsite power was between 1.2 x 10-1 and 1.1 x 10-7. Finally, the SPAR model calculated that the conditional core damage probability for a loss of offsite power initiated in the switchyard was 5.3 x 10-5. Using these values, under all scenarios evaluated by the analyst, the change in core damage frequency caused by the subject performance deficiency was below 1 x 10-6. Therefore, the finding was of very low safety significance (Green). The inspectors determined the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement the corrective action program with a low threshold for identifying materials improperly stored in the 500 KV switchyard [P.1(a)](Section 1R01). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Loss of Alternate Method of Decay Heat Removal Due to Reactor Water Clean Up Pumps Tripping on Low Suction Flow Signal Green. The inspectors reviewed a self-revealing finding for the licensees failure to identify that de-energizing non-safety electrical bus 13BD1 and 13BD2 would cause the reactor water clean-up pumps A and B to trip on a low suction flow signal. On April 24, 2012, the plant was shut down for refueling outage 18, the residual heat removal system B was in service, and the reactor water clean-up system was in standby mode as the alternate shutdown cooling system. In this configuration, the plant was in yellow risk due to having two available systems for decay heat removal. At 10:00 a.m., both reactor water clean-up pumps tripped on low pump suction flow, causing the plant to enter an unplanned orange risk configuration for only having one system available for decay heat removal. The licensee determined the reactor water pumps tripped while opening the feeder breaker for the 13BD1 and 13BD2 buses (breaker 152-1305) for scheduled maintenance. When breaker 152-1305 was opened, optical isolator AT12 caused the pump low suction flow trip control contacts to close, which initiated the low suction flow alarm and caused the pumps to trip. Immediate corrective actions included restoring reactor water clean-up as the alternative source of decay heat removal by closing breaker 152-1305 and re-energizing the 13BD1 and 13BD2 buses. The licensee entered Page 9 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 this issue into their corrective action program as Condition Reports CR-GGN-2012-06092 and CR-GGN-2012-06105. The finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and it affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a loss of a system safety function. The inspectors determined that the cause of this finding was a latent issue; therefore no cross-cutting aspect was assigned (Section 1R13). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions to Address Configuration Control of Previous Non-cited Violation Green. The inspectors identified a non-cited violation of very low safety significance of 10 CFR Appendix B, Criterion XVI, "Corrective Action," for failure to implement adequate corrective actions for a previous NRC-identified non-cited violation. The previous finding involved a failure to maintain configuration control of various systems in the plant. In response to the previous finding, the licensee performed an apparent cause evaluation and developed actions to address the causes and extent of condition. However, the inspector identified that the actions pertaining to the extent of condition were not properly implemented and, as a result, the deficiency identified by the inspector was not fully resolved. The licensee failed to identify brass compression fittings installed on drain tailpieces of the standby service water system instead of stainless steel fittings as required by design documents. Furthermore, the licensee failed to update applicable design drawings allowing sacrificial compression fittings to be installed. The licensee performed corrective actions to restore configuration control. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-04003, CR-GGN-2012-4180, and CR-GGN-2012-04233. The issue is more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, the issues identified by the inspector impacted the licensees ability to establish and maintain configuration control for equipment relied on for safe operation of the plant. The design control attribute of the Mitigating Systems Cornerstone and the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences were affected. Until the issues are fully resolved, the licensee continues to be vulnerable to gaps in their system configuration control. The finding was determined to be of very low safety significance (Green) using Attachment 4 to IMC 0609, "Significance Determination Process," because it did not result in an actual loss of safety function. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance associated with the resources component because the licensee did not provide adequate training of personnel so that the inappropriately installed fittings could be identified during system walkdowns [H.2(b)] (Section 1R08). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement a Surveillance Requirement to Assure that the Limiting Condition for Operation Will be Met Green. The inspectors identified a non-cited violation of 10 CFR Part 50.36, Technical Specifications, involving the failure to implement a surveillance requirement to assure that the limiting conditions for operation of the ultimate heat sink will be met. Technical Specifications requires two cooling towers and two cooling basins, with the volume of the two basins constituting the entire inventory of the ultimate heat sink. Therefore, an interconnecting siphon line is installed to transfer water between the two cooling tower basins. That siphon line has the safety-related function of ensuring the availability of enough cooling water to satisfy ultimate heat sink requirements. Technical Specification 3.7.1 includes Surveillance Requirement 3.7.1.1, which verifies the water level in each cooling tower basin every 24 hours, and Surveillance Requirement 3.7.1.2, which verifies each cooling tower fan every 31 days. However, the inspectors identified that Technical Specification 3.7.1 does not include a surveillance requirement to verify that the Page 10 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 interconnecting siphon line will perform its safety-related function. On May 20, 2012, the licensee performed an operability test for the siphon line and determined that it was operable. The licensee is currently performing a preventative maintenance task as a compensatory action to ensure operability of the siphon line until a license amendment can be submitted to the NRC that establishes a surveillance requirement. The licensee documented this violation in Condition Reports CR-GGN-2012-08257 and CR-GGN-2012-08537. The violation is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, without a surveillance requirement that verifies the interconnecting siphon line can perform its safety-related function, the licensee cannot ensure that sufficient cooling water is available following an accident. The inspectors evaluated the finding using Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings and determined that the finding was of very low safety significance (Green) because the finding was a design or qualification deficiency confirmed not to result in a loss of operability or function; did not represent a loss of safety system function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time; 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 in the human performance area associated with the resources component because the licensee did not ensure that equipment was adequate to assure nuclear safety, in that the licensee had recently reviewed documentation associated with a modification to the siphon line but failed to identify that operability of the UHS could not be established without a technical specification surveillance requirement to ensure operability of the siphon line [H.2(c)] (Section 1R19). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow a Post-Modification Test Procedure Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to follow a post-modification test procedure for the interconnecting siphon line between the two standby service water system cooling tower basins. Operability of the ultimate heat sink is based on a minimum water level in the two standby service water cooling tower basins, an operable interconnecting siphon between the basins, and four operable cooling tower fans (two per basin). At extended power uprate conditions, the configuration of the basins and the original siphon line would not support 30 days of operation of both trains of the standby service water system and the high pressure core spray service water systems without makeup, so the licensee performed a modification (EC 25649), which involved replacing the original siphon line with a new siphon line in order to transfer water from one basin to the other. On March 28, 2012, after completing the modification, the licensee performed post-modification testing to determine the piping friction loss coefficient of the modified siphon line and to evaluate its acceptability against the worst-case friction loss coefficient documented in EC 25649. The licensee deviated from the test procedure, as-written, and performed the test with an inadequate pressure gauge instead of the specified gauge. After inspectors challenged the validity of these test results, the licensee performed another test of the siphon line with a different method that did not require the use of a pressure gauge to measure the piping friction loss coefficient. The inspectors reviewed the subsequent test data and found the test results to be satisfactory. The licensee documented this concern in Condition Report CR-GGN-2012-05260. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the use of an unqualified gauge invalidated the test results, and a different test method had to be developed to determine the piping friction loss coefficient for the siphon line. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed to result in loss of operability or function; did not represent a loss of safety system function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time; 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 in the human performance area associated with work practices component because Page 11 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 licensee personnel proceeded in the face of uncertainty or unexpected circumstances. Specifically, the licensee proceeded with the test without verifying that the pressure gauge was suitable for the test conditions after observing unexpected measurements with the gauge [H.1(a)] (Section 1R19). Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Take Timely Corrective Actions Associated with Division 1 and 2 Standby Service Water Safety Related Cables that were Partially Submerged in Cable Manhole/Vault Green . The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take timely corrective actions to correct a condition adverse to quality associated with division 1 and 2 standby service water safety related cables that were partially submerged in a cable manhole/vault. The inspectors reviewed work order 52284535 and noted that the sump pump for manhole/vault MH-01, which contained standby service water cables for division 1 and 2, was determined to be non-functional on September 10, 2011. The inspectors determined that a work order to repair the non-functioning sump pump had been developed but that the work order had not yet been scheduled. During a subsequent inspection, manhole/vault MH-01 was found to contain approximately three feet of water, with water partially covering some of the safety related cables. The electricians immediately pumped manhole/vault MH-01 and wrote a condition report. The licensee repaired the sump pump the next week and declared it functional. The cables remained operable based on the results of meggar tests. The licensee entered this issue into their corrective action program as condition reports CR-GGN-2012-00503, 01324, and 01389. The finding is more than minor because it is associated with the equipment performance attribute of Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent either a loss of system safety function or an actual loss of safety function of a single train of one or more non-Technical Specification 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 in the area of human performance associated with the work practices component, in that the licensee personnel did not initiate a condition report as required by licensee procedure when the work order associated with sump pump testing of MH-01 determined that the sump pump was not functioning properly [H.4(b)] (Section 1R06). Inspection Report# : 2012002 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Condition Adverse to Fire Protection, in That the Licensee Failed to Adequately Provide Contingency Lighting in the Fire Brigade Dress Out Area While Normal Lighting was Inoperabl Green. The inspectors identified a Green non-cited violation of Facility Operating License Condition 2.C(41), for the failure to correct a condition adverse to fire protection. Specifically, the licensee failed to adequately provide contingency lighting in the fire brigade dress out area while normal lighting was inoperable due to maintenance on an associated breaker. The inadequate lighting delayed fire brigade response to a potential fire in the turbine building. Immediate corrective action included placing temporary lighting in the area. Normal lighting to the area was restored the next week. The licensee entered this issue into their corrective action program as condition report CR-GGN-2012-01488. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined from table 3b that issues related to performance of the fire brigade are not included in Appendix F and require NRC management review using Appendix M. Regional management review evaluated the overall impact of lighting issue Page 12 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 in the fire brigade dress out area and concluded that, while the fire protection defense-in-depth was affected by the performance deficiency, the overall defense-in-depth of the front-line systems was not impacted because of train separation and safe shutdown analysis at the site. Therefore the finding screened as having very low safety significance (Green) in accordance with Manual Chapter 0609, Appendix M. The inspectors determined the finding had a cross-cutting aspect in the area of human performance associated with the work control component, in that licensee personnel failed to ensure adequate job site conditions (lighting in the fire bridge dress out area) were in place prior to performance electrical maintenance in the turbine building [H.3(a)] (Section 40A3). Inspection Report# : 2012002 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: VIO Violation Inadequate Corrective Action for a Leak on the Division II Emergency Diesel Generator Lube Oil Sump Green. The team identified a Green cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a leak on the Division II emergency diesel generator lube oil sump. Despite identification of the leak in 2004, ineffective attempts to repair the leak and previous identification by the NRC in 2009, the licensee dispositioned the leak as accept as-is without a full understanding of the lube oil sump leak and potential consequences. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8880. The condition was discovered and documented by the licensee in 2004. This finding was initially determined by the NRC to be a minor violation in 2009. Paragraph F of Section 2.10 of the NRC Enforcement Manual states in part that where a licensee does not take corrective action for a minor violation, the matter should be considered more than minor and associated with a green inspection finding and dispositioned in a cited or noncited violation, as appropriate. This finding is now determined to be more than minor because if left uncorrected the failure to restore the lube oil sump for the Division II emergency diesel generator to design conditions would have the potential to lead to a more significant safety concern, specifically, the leak could worsen and potentially affect operability of the emergency diesel generator. Due to the licensees failure to restore compliance within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy. This finding affects the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate this problem such that the resolutions addressed the causes [P.1(c)]. (Section 4OA2.5d) Inspection Report# : 2011006 (pdf) Barrier Integrity Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Make Timely Corrective Actions to Repair the Degraded Auxiliary Building Water Intrusion Barrier Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to correct a condition adverse to quality in a timely manner. Specifically, the licensee failed to correct multiple degraded conditions associated with the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-10314. Corrective actions included generating Work Order 318398 and delegating funds to repair the water intrusion barrier at the next available opportunity. The finding is more than minor because if left uncorrected, the condition of a degraded auxiliary building water Page 13 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 intrusion barrier could lead to a more significant safety concern. Specifically, continued degradation of the water intrusion barrier could lead to the auxiliary building (secondary containment) being degraded such that the standby gas treatment system would not be able to achieve and maintain the design negative pressure of 1/4 inch water column within 120 seconds. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding had very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that the apparent cause of this finding was that the licensee had failed to classify the degraded water intrusion barrier as a condition adverse to quality that warranted correction in a timely manner. Therefore, the finding has a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly classify conditions adverse to quality [P.1(c)](Section 1R12). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Condition of the Auxiliary Building Water Intrusion Barrier Green. The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2), for the failure to monitor the performance of the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11740. Corrective actions included initiating Condition Report CR-GGN-2012-12286, in which the licensee concluded the degraded water intrusion barrier had experienced a Maintenance Rule Functional Failure and required further evaluation to determine if the barrier should be classified in 10 CFR 50.65 (a)(1). The finding is more than minor because if left uncorrected, the failure to monitor the performance of the auxiliary building water intrusion barrier in accordance with the maintenance rule program could lead to a more significant safety concern. Specifically, continued unmonitored degradation of the water intrusion barrier could compromise the integrity of the secondary containment function of the auxiliary building. Using Inspection Manual Chapter 0609, , Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding had a very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that the apparent cause of this finding was the licensee failed to recognize that the auxiliary building water intrusion barrier was scoped into their Maintenance Rule program with the monitoring criteria of zero occurrences of water intrusion barrier degradation. Therefore, the finding had a cross-cutting aspect in the human performance area, work practices component because the licensee failed to follow maintenance rule program procedures [H.4(b)](Section 1R12). Inspection Report# : 2012005 (pdf) Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Loss of Decay Heat Removal to the Spent Fuel Pool Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specifications 5.4.1(a), involving a loss of decay heat removal in the spent fuel pool due to station personnel failing to correctly follow operation of pool gate seal air supply procedure. On April 17, 2012, Grand Gulf Nuclear Station was preparing to drain the reactor cavity to reinstall the vessel head after the completion of refueling activities. In preparation, the upper containment pool to the reactor cavity gate was installed by General Electric-Hitachi technicians with Entergy oversight. Technicians were directed by procedure to verify that all supply isolation toggle valves to the gate seals were open and secured in place. However, technicians failed to complete this action correctly and the control room was informed that all prerequisites were completed and began the cavity drain down. The control room immediately noticed the fuel pool drain tank level was decreasing and attempted to makeup to the tank via the normal makeup valve. When the fuel pool drain tank level Page 14 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 reached 17 percent full, both fuel pool cooling and cleanup pumps tripped as expected, resulting in loss of decay heat removal to the spent fuel pool. The main control room entered the off-normal event procedure for inadequate decay heat removal, and they secured the drain down evolution. Approximately 47 minutes later, spent fuel pool cooling was re-established. During this event, the spent fuel pool temperature did not exceed the limits required by Technical Requirements Manual Section 6.7.4 (140°F). Short term corrective actions included restoring decay heat removal to the spent fuel pool and conducting a human performance review of the event. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-05756. The finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represented a loss of spent fuel pool cooling that would not preclude restoration of cooling to the spent fuel pool prior to pool boiling. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to use adequate self- and peer-checking techniques to ensure gate seals were properly inflated prior to cavity drain down [H.4(a)] (Section 1R20). Inspection Report# : 2012003 (pdf) Significance: Mar 23, 2012 Identified By: NRC Item Type: NCV NonCited Violation Modification of the Spent Fuel Pool without Prior NRC Approval SLIV. The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests and Experiments, when the licensee failed to obtain a license amendment prior to implementing a proposed change to the plant that required a change to Technical Specifications. The 10 CFR 50.59 evaluation performed by the licensee is dated January 24, 2001, thus it was performed under the requirements of the old rule based on the Entergy Operations letter dated March 5, 2001. In the 10 CFR 50.59 evaluation for the removal of Blackness Testing and the division of the spent fuel pool into two regions, the licensee determined that the modifications did not require a change to Technical Specifications. However, 10 CFR 50.36, Technical Specifications, Section 4, Design Features, requires that design features such as geometric arrangements, which, if altered or modified, would have a significant effect on safety, must be incorporated into Technical Specifications. The NRC considers that the establishment of two regional zones in the spent fuel pool, each having specific loading criteria to maintain keff less than 0.95, constitutes design features which, if altered or modified would have a significant effect on safety. Therefore, these design features should have been incorporated into the Technical Specifications. In a letter dated September 8, 2010, (ML102660403), the licensee submitted a power up-rate license amendment request. The NRC staff is currently reviewing the license request, which includes the licensees technical justification for the spent fuel pool changes described above. Based on preliminary review of the amendment request, the NRC staff has determined that an immediate safety concern does not exist. The licensee has entered this issue into their corrective action program as condition report CR-GGN-2012-01077. The finding is more than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, containment) protect the public from radionuclide releases caused by accidents or events. Inspectors performed a Phase 1 screening, in accordance with Inspection Manual Chapter 0609, , Phase 1 - Initial Screening and Characteristics of Findings, and determined that the finding was of very low safety significance (Green) because it did not result in the loss of cooling to the spent fuel pool, did not result from fuel handling errors that caused damage to fuel clad integrity, and it did not result in a loss of spent fuel pool inventory. This finding is a latent issue and is not indicative of current performance; therefore, no cross-cutting aspect was identified (Section 1R15). Inspection Report# : 2012002 (pdf) Emergency Preparedness Page 15 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 Occupational Radiation Safety Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Adequately Plan and Control Work Activities to Maintain ALARA Green. The inspector reviewed a self-revealing finding of very low safety significance because during the refueling outage 18 extended power upgrade, the licensee did not adequately plan and control work activities for the design and replacement of the new fuel pool cooling heat exchangers. Specifically, outage personnel did not perform adequate pre-outage walkdowns, which resulted in significant unplanned collective exposure. Actual collective dose and hours for Radiation Work Permit 2012-1086, Fuel Pool Cooling & Cleanup Heat Exchanger Replacement, was 23.9 person-rem and 12,237 RWP-hours, respectively. This is compared to the initial planned estimate of 3.74 person-rem and 1,905 RWP-hours. This finding and procedural concern was entered into the corrective action program as Condition Reports CR-GGNS-2012-09011 and CR-GGNS-2012-12398. The failure to appropriately use ALARA planning and controls procedures to prevent unplanned and unintended collective doses was a performance deficiency. This performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone attribute of Program and Process in that the failure to adequately implement ALARA procedures caused the collective radiation dose for the job activity to exceed the planned dose by more than 50 percent. In addition, this type of issue is addressed in Example 6.j of IMC 0612, Appendix E, Examples of Minor Issues. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this performance deficiency to be a finding of very low safety significance because although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 240 person-rem. This finding has a cross-cutting aspect in the human performance area, work control component, because the licensee failed to evaluate the impact of work scope change on human performance and interdepartmental communication and coordination prior to commencing work activities. Specifically, there was inappropriate coordination and communication of work activities between work groups [H.3(b)](Section 2RS02). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Follow the Radiation Work Permit Requirements During Reactor Cavity High Water Operations Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1 for failure to comply with radiological exposure controls specified in Radiation Work Permit 2012-1402, Refuel Floor High Water Activities. Specifically, radiation exposure controls in the RWP required the licensee to verify that fuel pool cleanup (demineralizers) was in-service, and if dose rates increased by more than 0.2 millirem/hour, change the resins. During reactor cavity operations, both fuel pool demineralizer trains were inoperable at least 25 days. In addition, the dryer separator pool and reactor cavity were isolated from the fuel pool clean up system. Consequently, general area radiation levels on the reactor cavity floor increased from 0.4 millirem/hour to 6.0 millirem/hour. The actual collective dose and hours for the work activity was 8.24 person-rem and 9,000 RWP-hours, respectively. This is compared to the planned initial estimate of 4.60 person-rem and 6,987 RWP-hours. This Radiation Work Permint and procedure violation was documented in the licensees corrective action program as Condition Reports CR-GGNS-2012-04288 and CR-GGNS-2012-12401. The licensees failure to comply with the RWP to prevent unplanned and unintended collective doses was a performance deficiency. This performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone attribute of Program and Process in that the failure to adequately implement ALARA procedures caused the collective radiation dose for the job activity to exceed the planned dose by more than 50 percent. In addition, this type of issue is addressed in Example 6.i of IMC 0612, Appendix E, Examples of Minor Issues. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this performance deficiency to be a non-cited violation of very low safety significance because although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 240 person-rem. The violation involved a cross-cutting aspect in the human performance area, work control component, because the licensee did not appropriately coordinate work activities by incorporating actions to address the need for work groups to communicate and coordinate with each other during activities in which interdepartmental coordination was Page 16 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 necessary to assure human performance [H.3(b)](Section 2RS02). Inspection Report# : 2012005 (pdf) Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary. Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Use of Waivers to Allow Workers to Exceed the Minimum Day Off Rule Green. The inspectors identified a non-cited violation of 10 CFR 26, Subpart I, Managing Fatigue, Subsection 207, Waivers and Exceptions, when the licensee inappropriately used waivers to allow workers to exceed the minimum day off rule. While reviewing condition reports, the inspectors noted the use of work hour waivers for a large number of staff. The circumstances for the use of waivers were the refueling outage lasting more than 60 days, contract expiration leading to 14 layoffs, and the loss of 4 workers via voluntary resignation. Due to these circumstances, work hours and fatigue of waivered individuals would have to be assessed daily. The assessment is required because the work hour limit of these individuals exceeded the minimum day off rule, therefore requiring daily monitoring until the end of the cycle. The waivered individuals averaged two days off per six-week period compared to the required three days off. Title 10 CFR 26.207 (a)(2) allows the granting of waivers only to address circumstances that could not have been reasonably controlled. The inspectors determined that the licensee was aware of the circumstances of an extended refueling outage and contract renewal deadline well in advance of the need to grant waivers, and a reasonable amount of time was available for the licensee to develop and execute contingency plans to negate the need to use waivers. Corrective actions included initiating assessments and waivers for exceeding minimum days off requirements for shift personnel for the six-week period ending May 27, 2012, and returning to the normal on-line work schedule in which adequate manpower is available to meet the requirements of the rule. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-7348. The finding is more than minor because it is associated with the access authorization attribute of the Security Cornerstone, and affected the cornerstone objective to provide assurance that the licensees security system and material control and accounting program use a defense in-depth approach and can protect against (1) the design basis threat of radiological sabotage from external and internal threats, and (2) the theft or loss of radiological materials. Using the Inspection Manual Chapter 0609, Appendix E, Baseline Security Significance Determination Process for Power Reactors, Figures 5 and 6, the finding was determined to have very low security significance because the calculated point total did not exceed the threshold value for a Green non-cited violation. The cumulative total for this finding was zero points, which was calculated by factoring the one impact area (vital areas) against Tier III Element 08.02.08, security force work hours, of the access authorization attribute, which resulted in a total of zero points within this attribute. The finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision making component in that the licensee failed to use conservative assumptions in developing staff schedules for the duration of refueling outage 18 and for allowing an employment contract to expire Page 17 of 18

4Q/2012 Inspection Findings - Grand Gulf 1 that led to 14 individuals being laid off without realizing the impact these decisions would have on the licensees ability to meet the requirements of the rule [H.1(b)] (Section 1R20). Inspection Report# : 2012003 (pdf) Last modified : February 28, 2013 Page 18 of 18

1Q/2013 Inspection Findings - Grand Gulf 1 Grand Gulf 1 1Q/2013 Plant Inspection Findings Initiating Events Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Risk Significances and Develop Action Plans to Address Equipment Identified During Extent of Condition Review for a Post Scram Root Cause Analysis Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to follow procedure EN-LI-118, Root Cause Evaluation Process, Revision 18, in that they failed to evaluate the risk significances and develop action plans to address equipment identified during their extent-of-condition review for a post-scram root cause analysis. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11950. The immediate corrective actions included assigning corrective actions for operations personnel to properly evaluate the risk significance of the identified components and perform appropriate corrective actions to correct the degraded conditions. The licensees failure to properly determine risk significance and associated action plans to correct degraded equipment that could challenge safe plant operation is a performance deficiency. The performance deficiency is more than minor and is therefore a finding because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to take corrective actions to correct degraded equipment has the potential to lead to initiating events resulting in plant transients. Using NRC Inspection Manual Chapter 0609, , "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the apparent cause of this finding was that when operations management directed operators to identify the degraded equipment, they did not encourage those operators to comply with Procedure EN-LI-118. Therefore, the finding has a cross-cutting aspect in the human performance area, work practices component because the licensee did not define and effectively communicate expectations regarding procedural compliance. [H.4 (b)] (Section 4OA3). Inspection Report# : 2012005 (pdf) Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure of Hot Work Fire Watch to Follow Procedural Requirements The inspectors reviewed a self-revealing non-cited violation of Technical Specifications 5.4.1(a), for failure of the hot-work fire watch to follow procedural requirements, which resulted in a fire in main condenser A. On April 11, 2012, at 6:11 p.m., hot-work was in progress inside the condenser A in the upper southeast corner at 150 foot elevation. Cutting was being performed by contract boilermakers using an oxy-acetylene torch, with ventilation exhaust and supply provided by nearby HEPA hoses. The torch cutting operation produced hot slag, which exited the barrier provided by the fire blankets and ignited the nearby HEPA hoses, air conditioning hoses, and eventually the acetylene hoses. Contract pipefitters in the area were able to extinguish the fire. The main control room was informed Page 1 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 of the fire inside condenser A and dispatched the fire brigade to the scene. The operations shift manager declared a notice of unusual event at 6:26 p.m. due to a fire in the protected area lasting longer than 15 minutes. Members of the fire brigade entered the condenser bay at 6:42 p.m. and reported to the control room there was no fire present, only smoke. The notice of unusual event was exited at 7:00 p.m. Short term corrective actions included site management placing a stop work order on all hot-work until a complete investigation of the event could be performed. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-05418. The finding is more than minor because it is associated with the protection against external factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors reviewed Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," that states in the Assumptions and Limitations section, The Fire Protection SDP focuses on risks due to degraded conditions of the fire protection program during full power operation of a nuclear power plant. This tool does not address the potential risk significance of fire protection inspection findings in the context of other modes of plant operation (i.e., low power or shutdown). Therefore, the senior reactor analyst evaluated the finding in accordance with Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for both PWRs and BWRs. The finding did not require a quantitative assessment because adequate mitigating equipment remained available; the finding did not increase the likelihood of a loss of reactor coolant system inventory; the finding did not degrade the ability to terminate a leak path or add reactor coolant system inventory; and the finding did not degrade the ability to recover decay heat removal if lost. Therefore, the finding screened as Green, having very low safety significance. The inspectors determined that the apparent cause of this finding was that site management did not ensure that hot-work supervisors were engaged in ensuring compliance with procedural requirements. This finding had a cross-cutting aspect in the area of human performance associated with work practices component because the licensee failed to ensure supervisory oversight of hot-work activities is performed within procedural requirements such that nuclear safety is supported [H.4(c)] (Section 40A3) Inspection Report# : 2012003 (pdf) Mitigating Systems Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Adequate Procedure Instructions to Perform Preventive Maintenance Requiring The Periodic Replacement of the Control Relays in the GE Magne Blast Circuit Breakers Green. The inspectors reviewed a self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to complete preventive maintenance tasks on the high pressure core spray division III diesel generator output breaker in accordance with the corresponding preventive maintenance task template. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2012-07992. The immediate corrective actions included replacing the failed control relay and restoring operability to the division III diesel generator. The long term corrective actions included revising breaker refurbishment/replacement procedure with directions to replace the control relay and change the procedure frequency to every 10 years versus every 12 years. The inspectors determined that this performance deficiency was more than minor and is therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this failed control relay caused the subject breaker to fail to close during the division III diesel generator monthly surveillance on June 5, 2012. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," to determine that the issue Page 2 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 affected the Mitigating System Cornerstone. Because the finding pertained only to a degraded condition while the plant was shutdown, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Checklist 8, Cold Shutdown or Refueling Operation - Time to Boil > 2 Hours: RCS Level < 23 Above Top of Flange, to determine that the finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add RCS inventory when needed; did not significantly degrade the licensees ability to recover decay heat removal if lost; and did not affect the safety/relief valves (Green). The inspectors determined that the cause of this finding was a latent issue that is not reflective of current performance, therefore no cross-cutting aspect was identified. (Section 1R20.b). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Gain Settings on APRM and LPRM Instruments in Accordance with Design Requirements Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish the gain settings used on the power range neutron monitoring system in accordance with design requirements. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2013-00177. The immediate corrective actions included adjusting gain settings for their average power range monitor (APRM) instruments to indicate actual core thermal power as determined by the heat balance. In additioin, the licensee revised their neutron monitoring procedure to set the initial gains for the average power range monitor to the maximum value to maintain conservative power indication during future startups. They also changed their local power range monitor replacement procedure to use the vendor specified initial gain setting of 3.692 prior to startup. The finding was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect gain settings caused a violation of technical specification 3.0.4 by rendering the APRM Neutron Flux High - Setdown scram function and the Neutron Flux - Upscale, Startup control rod block function inoperable prior to entry into Mode 2. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had very low safety significance (Green) because although the finding affected a single reactor protection system trip signal to initiate a reactor scram, it did not affect the function of other redundant trips or diverse methods of reactor shutdown, did not involve control manipulations that unintentionally added positive reactivity, and did not result in a mismanagement of reactivity by operators. Because the performance deficiency occurred in the past and is not reflective of current licensee performance, this finding was not assigned a cross-cutting aspect. (Section 4OA3). Inspection Report# : 2012005 (pdf) Significance: Sep 21, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Operability Determinations Green. The inspectors identified two examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure EN-OP-104, Operability Determinations. Specifically, for Condition Report CR-GGN-2012-09690, which documents an oil leak on the standby liquid control pump B, and for Condition Report CR-GGN-2012-09889, which documents degraded bolts on a flanged connection on standby service water B piping, the licensee failed to validate that operability evaluations completed for prior non-conforming conditions bounded the Page 3 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 conditions documented in the new condition reports. As immediate corrective actions, the licensee re-performed the evaluations and established an adequate basis for operability for the conditions described in the two condition reports listed above. The licensee entered this issue into their corrective action program as CR-GGN-2012-09735 and CR-GGN-2012-10664. The finding was more than minor because if left uncorrected, not performing operability determinations in accordance with procedure could lead to a more significant safety concern. Specifically, if a condition renders a safety related system inoperable and because of this performance deficiency the licensee incorrectly determines that the system is operable, then this performance deficiency could result in a safety related system remaining inoperable for a long period of time. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because although it affected the design or qualification of a mitigating system, the system maintained its operability. The finding had a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly evaluate for operability conditions adverse to quality [P.1(c)] (Section 1R15). Inspection Report# : 2012004 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of 4160 Vac Circuit Breakers for As-Found Tests Green. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to July 27, 2012, the licensees preventive maintenance Procedures 07-S-12-41, 07-S-12-42, and 07-S-12-61 failed to assure that the 4160 Vac circuit breakers would perform satisfactorily in service when the licensee performed maintenance prior to completing as-found tests to verify past operability of the circuit breakers. This finding has been entered into licensees corrective action program as Condition Reports CR-GGN-2012-09035 and CR- GGN-2012-9103. The team determined that failure to establish a test program which ensures that test and maintenance procedures associated with safety-related 4160 Vac circuit breakers would perform satisfactorily in service was a performance deficiency. This finding was more than minor because, if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to perform as-found tests prior to performing maintenance in preventive maintenance procedures was a significant programmatic deficiency which could cause unacceptable conditions to go undetected. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee failed to ensure that test and maintenance procedures were complete, accurate, and up-to-date to assure nuclear safety. [H.2(c)] (1R21.2.1) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Page 4 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 Item Type: NCV NonCited Violation Failure to Establish a Testing Program for Safety Related 125 Vdc Circuit Breakers Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to July 27, 2012, the licensee failed to establish a test program for 125 Vdc safety related molded case circuit breakers incorporating the requirements of IEEE 308, to ensure the breakers would not degrade and would perform satisfactorily in service. The finding was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-09030 and CR-GGN-2012-09175. The team determined that the failure to establish a testing program incorporating the requirements of IEEE 308 was a performance deficiency. The finding was more than minor, because if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to establish a testing program was a significant programmatic deficiency that would lead to missed opportunities to detect potential common cause failures from degradation of performance in more than one redundant safety division. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component; because the licensee failed to thoroughly evaluate problems such that resolutions address cause and extent of condition. Specifically, the licensee failed to thoroughly evaluate the extent of condition associated with previously identified NRC violation involving the failure to test 480 Vac molded case circuit breakers identified during the 2009 component design basis inspection. [P.1(c)] (1R21.2.2) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for a Change to Credible Passive Failures in the Standby Service Water System Severity Level IV. The team identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests and Experiments which states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if this activity would; result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report (as updated). Specifically, on August 18, 1987, the licensee implemented a change to the updated safety analysis report which limited credible passive failures in the standby service water system to pump and valve seal leakage without obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR GGN 2012 09267. The team determined that the licensees failure to receive prior NRC approval for changes in licensed activities regarding single passive failure criteria for the standby service water system was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding had the ability to impact the regulatory process. The performance deficiency was more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation. In accordance with the NRC Enforcement Manual, risk insights from the Inspection Manual Chapter 0609, Significance Determination Process, are used in determining the significance of 10 CFR 50.59 violations. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Page 5 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because, although the standby service water system could not provide 30 days of decay heat removal without operator action to provide makeup water to the system, it would have been able to complete its 24-hour risk significant mission time. Since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. The finding does not have a crosscutting aspect because the most significant contributor to the finding does not reflect current licensee performance. (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Enter an NRC Violation Regarding the Standby Service Water System into the Corrective Action Program Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, on July 12, 2012, the NRC informed the licensee of a violation of 10 CFR 50.59 requirements, but the licensee failed to promptly identify this as an adverse condition and enter this condition into their corrective action program until July 19, 2012. The finding was entered into the licensees corrective action program as CR-GGN-2012-10075. The team determined that the licensees failure to promptly enter the NRC violation as condition adverse to quality into the corrective action program was a performance deficiency. This finding was more than minor because it adversely affected the design control attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to promptly document a violation of 10 CFR 50.59, which delayed an operability evaluation that ultimately determined that compensatory measures were required to ensure that the standby service water system could perform its specified safety function for its entire mission time. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because, although the standby service water system could not provide 30 days of decay heat removal without operator action to provide makeup water to the system, it would have been able to complete its 24-hour risk significant mission time. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee Page 6 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 failed to ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated, and that actions are taken to address safety issues, in a timely manner, commensurate with their safety significance. Specifically, the licensee did not implement a corrective action program with a low threshold for identifying issues completely, accurately, and in a timely manner commensurate with their safety significance. [P.1(a)] (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Operability Determination Process Procedure Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings which states, in part, that Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, from July 19, 2012, to July 29, 2012, the licensee failed correctly evaluate the operability of the standby service water system with a degraded or nonconforming condition and failed to document a sound basis for a reasonable expectation of operability of the standby service water system as required by Procedure EN-OP-104, Operability Determination Process. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2012-09356. The team determined that the failure to implement the requirements of the operability determination process procedure was a performance deficiency. The finding was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the standby service water system was incapable of performing its specified safety function for the entire 30-day mission time without compensatory measures. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because the standby service water system could would have been able to complete its 24-hour risk significant mission time although it could not provide 30 days of decay heat removal without operator action to provide makeup water to the system. This finding had a crosscutting aspect in the area of human performance, decision making component, because the licensee did not make decisions that demonstrated that nuclear safety was an overriding priority. Specifically, the licensee did not make safety significant decisions using a systematic process to ensure safety is maintained. [H.1(a)] (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Test and Inspection Requirements for 4160 Vac Circuit Breakers into Preventive Page 7 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 Maintenance Procedures Green. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design document. Specifically, prior to July 27, 2012, the licensees safety-related 4160 Vac circuit breaker preventive maintenance Procedures 07-S-12-41, 07-S-12-42, and 07 S 12-61 failed to incorporate inspection and test requirements for minimum voltage tests, reduced voltage tests, and inspection of auxiliary switch relay contacts as established in the licensees circuit breaker maintenance program. This condition was entered into the licensees corrective action program as Condition Reports CR GGN 2012-08885 and CR-GGN-2012-09111. The team determined that the failure to incorporate required tests and inspections into preventive maintenance procedures for safety related 4160 Vac circuit breakers was a performance deficiency. This finding was more than minor because, if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to incorporate the testing, cleaning, and inspection requirements into preventive maintenance procedures were a significant programmatic deficiency which could cause unacceptable conditions to go undetected. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of problem identification and resolution, operating experience component, because the licensee failed to use operating experience information, including vendor recommendations and internally generated lessons learned, to support plant safety. Specifically, the licensee did not implement and institutionalize operating experience through changes to processes, procedures, equipment, and training programs. [P.2(b)] (1R21.2.4) Inspection Report# : 2012008 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: FIN Finding Failure to Ensure Materials are Stored Properly in the 500 KV Switchyard The inspectors identified a finding for the licensees failure to ensure that materials or equipment were not stored under energized lines or near energized equipment in accordance with station procedures. On May 21, 2012, the inspectors were performing a grid stability inspection and toured the 500 KV switchyard with the system switchyard engineer. During the tour, the inspectors identified numerous cylindrical shaped items stored under a 500 KV power line, which posed a missile hazard to the offsite source of power. The licensee determined that the items in question were bushing sleeves that were left in the switchyard following 500 KV breaker maintenance. The inspectors researched station procedures and determined that the cylindrical items stored under the energized 500 KV power line did not meet procedure requirements for the storage of materials and equipment. Immediate corrective actions included having the items removed from the switchyard. The licensee entered this issue into their corrective action program as Condition Report CR-GGNS-2012-07362. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigation Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors reviewed Manual Chapter 0609, Attachment A, Phase 1 - Initial Screening and Characterization of Findings. Attachment A, Table 4.a, states that a Phase 3 is required if the finding is potentially risk significant due to external initiating event core damage accident sequences. The inspectors determined that the failure to properly store the bushing sleeves in the switchyard could have resulted in a loss of offsite power during a severe weather initiating event. Therefore, the senior reactor analyst evaluated the finding to determine its significance using hand calculations and the site-specific SPAR model. The analyst determined that the probability of having straight-line winds or winds generated by hurricanes or tornados that were strong enough to throw the bushing sleeves into switchyard electrical Page 8 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 equipment was between 2.5 x 10-1 and 2.0 x 10-2 /year. The analyst also determined that the conditional probability that bushing sleeves thrown by winds would result in a loss of offsite power was between 1.2 x 10-1 and 1.1 x 10-7. Finally, the SPAR model calculated that the conditional core damage probability for a loss of offsite power initiated in the switchyard was 5.3 x 10-5. Using these values, under all scenarios evaluated by the analyst, the change in core damage frequency caused by the subject performance deficiency was below 1 x 10-6. Therefore, the finding was of very low safety significance (Green). The inspectors determined the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement the corrective action program with a low threshold for identifying materials improperly stored in the 500 KV switchyard [P.1(a)](Section 1R01). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: FIN Finding Loss of Alternate Method of Decay Heat Removal Due to Reactor Water Clean Up Pumps Tripping on Low Suction Flow Signal Green. The inspectors reviewed a self-revealing finding for the licensees failure to identify that de-energizing non-safety electrical bus 13BD1 and 13BD2 would cause the reactor water clean-up pumps A and B to trip on a low suction flow signal. On April 24, 2012, the plant was shut down for refueling outage 18, the residual heat removal system B was in service, and the reactor water clean-up system was in standby mode as the alternate shutdown cooling system. In this configuration, the plant was in yellow risk due to having two available systems for decay heat removal. At 10:00 a.m., both reactor water clean-up pumps tripped on low pump suction flow, causing the plant to enter an unplanned orange risk configuration for only having one system available for decay heat removal. The licensee determined the reactor water pumps tripped while opening the feeder breaker for the 13BD1 and 13BD2 buses (breaker 152-1305) for scheduled maintenance. When breaker 152-1305 was opened, optical isolator AT12 caused the pump low suction flow trip control contacts to close, which initiated the low suction flow alarm and caused the pumps to trip. Immediate corrective actions included restoring reactor water clean-up as the alternative source of decay heat removal by closing breaker 152-1305 and re-energizing the 13BD1 and 13BD2 buses. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2012-06092 and CR-GGN-2012-06105. The finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and it affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a loss of a system safety function. The inspectors determined that the cause of this finding was a latent issue; therefore no cross-cutting aspect was assigned (Section 1R13). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions to Address Configuration Control of Previous Non-cited Violation Green. The inspectors identified a non-cited violation of very low safety significance of 10 CFR Appendix B, Criterion XVI, "Corrective Action," for failure to implement adequate corrective actions for a previous NRC-identified non-cited violation. The previous finding involved a failure to maintain configuration control of various systems in the plant. In response to the previous finding, the licensee performed an apparent cause evaluation and developed actions to address the causes and extent of condition. However, the inspector identified that the actions pertaining to the extent of condition were not properly implemented and, as a result, the deficiency identified by the Page 9 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 inspector was not fully resolved. The licensee failed to identify brass compression fittings installed on drain tailpieces of the standby service water system instead of stainless steel fittings as required by design documents. Furthermore, the licensee failed to update applicable design drawings allowing sacrificial compression fittings to be installed. The licensee performed corrective actions to restore configuration control. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-04003, CR-GGN-2012-4180, and CR-GGN-2012-04233. The issue is more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, the issues identified by the inspector impacted the licensees ability to establish and maintain configuration control for equipment relied on for safe operation of the plant. The design control attribute of the Mitigating Systems Cornerstone and the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences were affected. Until the issues are fully resolved, the licensee continues to be vulnerable to gaps in their system configuration control. The finding was determined to be of very low safety significance (Green) using Attachment 4 to IMC 0609, "Significance Determination Process," because it did not result in an actual loss of safety function. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance associated with the resources component because the licensee did not provide adequate training of personnel so that the inappropriately installed fittings could be identified during system walkdowns [H.2(b)] (Section 1R08). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement a Surveillance Requirement to Assure that the Limiting Condition for Operation Will be Met Green. The inspectors identified a non-cited violation of 10 CFR Part 50.36, Technical Specifications, involving the failure to implement a surveillance requirement to assure that the limiting conditions for operation of the ultimate heat sink will be met. Technical Specifications requires two cooling towers and two cooling basins, with the volume of the two basins constituting the entire inventory of the ultimate heat sink. Therefore, an interconnecting siphon line is installed to transfer water between the two cooling tower basins. That siphon line has the safety-related function of ensuring the availability of enough cooling water to satisfy ultimate heat sink requirements. Technical Specification 3.7.1 includes Surveillance Requirement 3.7.1.1, which verifies the water level in each cooling tower basin every 24 hours, and Surveillance Requirement 3.7.1.2, which verifies each cooling tower fan every 31 days. However, the inspectors identified that Technical Specification 3.7.1 does not include a surveillance requirement to verify that the interconnecting siphon line will perform its safety-related function. On May 20, 2012, the licensee performed an operability test for the siphon line and determined that it was operable. The licensee is currently performing a preventative maintenance task as a compensatory action to ensure operability of the siphon line until a license amendment can be submitted to the NRC that establishes a surveillance requirement. The licensee documented this violation in Condition Reports CR-GGN-2012-08257 and CR-GGN-2012-08537. The violation is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, without a surveillance requirement that verifies the interconnecting siphon line can perform its safety-related function, the licensee cannot ensure that sufficient cooling water is available following an accident. The inspectors evaluated the finding using Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings and determined that the finding was of very low safety significance (Green) because the finding was a design or qualification deficiency confirmed not to result in a loss of operability or function; did not represent a loss of safety system function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time; 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 in the human performance area associated with the resources component because the licensee did not ensure that equipment was adequate to assure Page 10 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 nuclear safety, in that the licensee had recently reviewed documentation associated with a modification to the siphon line but failed to identify that operability of the UHS could not be established without a technical specification surveillance requirement to ensure operability of the siphon line [H.2(c)] (Section 1R19). Inspection Report# : 2012003 (pdf) Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow a Post-Modification Test Procedure Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to follow a post-modification test procedure for the interconnecting siphon line between the two standby service water system cooling tower basins. Operability of the ultimate heat sink is based on a minimum water level in the two standby service water cooling tower basins, an operable interconnecting siphon between the basins, and four operable cooling tower fans (two per basin). At extended power uprate conditions, the configuration of the basins and the original siphon line would not support 30 days of operation of both trains of the standby service water system and the high pressure core spray service water systems without makeup, so the licensee performed a modification (EC 25649), which involved replacing the original siphon line with a new siphon line in order to transfer water from one basin to the other. On March 28, 2012, after completing the modification, the licensee performed post-modification testing to determine the piping friction loss coefficient of the modified siphon line and to evaluate its acceptability against the worst-case friction loss coefficient documented in EC 25649. The licensee deviated from the test procedure, as-written, and performed the test with an inadequate pressure gauge instead of the specified gauge. After inspectors challenged the validity of these test results, the licensee performed another test of the siphon line with a different method that did not require the use of a pressure gauge to measure the piping friction loss coefficient. The inspectors reviewed the subsequent test data and found the test results to be satisfactory. The licensee documented this concern in Condition Report CR-GGN-2012-05260. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the use of an unqualified gauge invalidated the test results, and a different test method had to be developed to determine the piping friction loss coefficient for the siphon line. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed to result in loss of operability or function; did not represent a loss of safety system function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time; 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 in the human performance area associated with work practices component because licensee personnel proceeded in the face of uncertainty or unexpected circumstances. Specifically, the licensee proceeded with the test without verifying that the pressure gauge was suitable for the test conditions after observing unexpected measurements with the gauge [H.1(a)] (Section 1R19). Inspection Report# : 2012003 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: VIO Violation Inadequate Corrective Action for a Leak on the Division II Emergency Diesel Generator Lube Oil Sump Green. The team identified a Green cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a leak on the Division II emergency diesel generator lube oil sump. Page 11 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 Despite identification of the leak in 2004, ineffective attempts to repair the leak and previous identification by the NRC in 2009, the licensee dispositioned the leak as accept as-is without a full understanding of the lube oil sump leak and potential consequences. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8880. The condition was discovered and documented by the licensee in 2004. This finding was initially determined by the NRC to be a minor violation in 2009. Paragraph F of Section 2.10 of the NRC Enforcement Manual states in part that where a licensee does not take corrective action for a minor violation, the matter should be considered more than minor and associated with a green inspection finding and dispositioned in a cited or noncited violation, as appropriate. This finding is now determined to be more than minor because if left uncorrected the failure to restore the lube oil sump for the Division II emergency diesel generator to design conditions would have the potential to lead to a more significant safety concern, specifically, the leak could worsen and potentially affect operability of the emergency diesel generator. Due to the licensees failure to restore compliance within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy. This finding affects the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate this problem such that the resolutions addressed the causes [P.1(c)]. (Section 4OA2.5d) Inspection Report# : 2011006 (pdf) Barrier Integrity Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Make Timely Corrective Actions to Repair the Degraded Auxiliary Building Water Intrusion Barrier Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to correct a condition adverse to quality in a timely manner. Specifically, the licensee failed to correct multiple degraded conditions associated with the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-10314. Corrective actions included generating Work Order 318398 and delegating funds to repair the water intrusion barrier at the next available opportunity. The finding is more than minor because if left uncorrected, the condition of a degraded auxiliary building water intrusion barrier could lead to a more significant safety concern. Specifically, continued degradation of the water intrusion barrier could lead to the auxiliary building (secondary containment) being degraded such that the standby gas treatment system would not be able to achieve and maintain the design negative pressure of 1/4 inch water column within 120 seconds. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding had very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that the apparent cause of this finding was that the licensee had failed to classify the degraded water intrusion barrier as a condition adverse to quality that warranted correction in a timely manner. Page 12 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 Therefore, the finding has a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly classify conditions adverse to quality [P.1(c)](Section 1R12). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Condition of the Auxiliary Building Water Intrusion Barrier Green. The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2), for the failure to evaluate the condition of the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11740. Corrective actions included initiating Condition Report CR-GGN-2012-12286, in which the licensee concluded the degraded water intrusion barrier had experienced a Maintenance Rule Functional Failure and required further evaluation to determine if the barrier should be classified in 10 CFR 50.65 (a) (1). The finding is more than minor because if left uncorrected, the failure to adequately evaluate the condition of the auxiliary building water intrusion barrier in accordance with the maintenance rule program could lead to a more significant safety concern. Specifically, continued inadequate evaluation of the water intrusion barrier could compromise the integrity of the secondary containment function of the auxiliary building. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that this finding is a latent issue; therefore no cross cutting aspect was assigned (Section 1R12). Inspection Report# : 2012005 (pdf) Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Loss of Decay Heat Removal to the Spent Fuel Pool Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specifications 5.4.1(a), involving a loss of decay heat removal in the spent fuel pool due to station personnel failing to correctly follow operation of pool gate seal air supply procedure. On April 17, 2012, Grand Gulf Nuclear Station was preparing to drain the reactor cavity to reinstall the vessel head after the completion of refueling activities. In preparation, the upper containment pool to the reactor cavity gate was installed by General Electric-Hitachi technicians with Entergy oversight. Technicians were directed by procedure to verify that all supply isolation toggle valves to the gate seals were open and secured in place. However, technicians failed to complete this action correctly and the control room was informed that all prerequisites were completed and began the cavity drain down. The control room immediately noticed the fuel pool drain tank level was decreasing and attempted to makeup to the tank via the normal makeup valve. When the fuel pool drain tank level reached 17 percent full, both fuel pool cooling and cleanup pumps tripped as expected, resulting in loss of decay heat removal to the spent fuel pool. The main control room entered the off-normal event procedure for inadequate decay heat removal, and they secured the drain down evolution. Approximately 47 minutes later, spent fuel pool cooling was re-established. During this event, the spent fuel pool temperature did not exceed the limits required by Technical Requirements Manual Section 6.7.4 (140°F). Short term corrective actions included restoring decay heat removal to the spent fuel pool and conducting a human performance review of the event. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-05756. The finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity Page 13 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 Cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represented a loss of spent fuel pool cooling that would not preclude restoration of cooling to the spent fuel pool prior to pool boiling. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to use adequate self- and peer-checking techniques to ensure gate seals were properly inflated prior to cavity drain down [H.4(a)] (Section 1R20). Inspection Report# : 2012003 (pdf) Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Adequately Plan and Control Work Activities to Maintain ALARA Green. The inspector reviewed a self-revealing finding of very low safety significance because during the refueling outage 18 extended power upgrade, the licensee did not adequately plan and control work activities for the design and replacement of the new fuel pool cooling heat exchangers. Specifically, outage personnel did not perform adequate pre-outage walkdowns, which resulted in significant unplanned collective exposure. Actual collective dose and hours for Radiation Work Permit 2012-1086, Fuel Pool Cooling & Cleanup Heat Exchanger Replacement, was 23.9 person-rem and 12,237 RWP-hours, respectively. This is compared to the initial planned estimate of 3.74 person-rem and 1,905 RWP-hours. This finding and procedural concern was entered into the corrective action program as Condition Reports CR-GGNS-2012-09011 and CR-GGNS-2012-12398. The failure to appropriately use ALARA planning and controls procedures to prevent unplanned and unintended collective doses was a performance deficiency. This performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone attribute of Program and Process in that the failure to adequately implement ALARA procedures caused the collective radiation dose for the job activity to exceed the planned dose by more than 50 percent. In addition, this type of issue is addressed in Example 6.j of IMC 0612, Appendix E, Examples of Minor Issues. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this performance deficiency to be a finding of very low safety significance because although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 240 person-rem. This finding has a cross-cutting aspect in the human performance area, work control component, because the licensee failed to evaluate the impact of work scope change on human performance and interdepartmental communication and coordination prior to commencing work activities. Specifically, there was inappropriate coordination and communication of work activities between work groups [H.3(b)](Section 2RS02). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Follow the Radiation Work Permit Requirements During Reactor Cavity High Water Operations Page 14 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1 for failure to comply with radiological exposure controls specified in Radiation Work Permit 2012-1402, Refuel Floor High Water Activities. Specifically, radiation exposure controls in the RWP required the licensee to verify that fuel pool cleanup (demineralizers) was in-service, and if dose rates increased by more than 0.2 millirem/hour, change the resins. During reactor cavity operations, both fuel pool demineralizer trains were inoperable at least 25 days. In addition, the dryer separator pool and reactor cavity were isolated from the fuel pool clean up system. Consequently, general area radiation levels on the reactor cavity floor increased from 0.4 millirem/hour to 6.0 millirem/hour. The actual collective dose and hours for the work activity was 8.24 person-rem and 9,000 RWP-hours, respectively. This is compared to the planned initial estimate of 4.60 person-rem and 6,987 RWP-hours. This Radiation Work Permint and procedure violation was documented in the licensees corrective action program as Condition Reports CR-GGNS-2012-04288 and CR-GGNS-2012-12401. The licensees failure to comply with the RWP to prevent unplanned and unintended collective doses was a performance deficiency. This performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone attribute of Program and Process in that the failure to adequately implement ALARA procedures caused the collective radiation dose for the job activity to exceed the planned dose by more than 50 percent. In addition, this type of issue is addressed in Example 6.i of IMC 0612, Appendix E, Examples of Minor Issues. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this performance deficiency to be a non-cited violation of very low safety significance because although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 240 person-rem. The violation involved a cross-cutting aspect in the human performance area, work control component, because the licensee did not appropriately coordinate work activities by incorporating actions to address the need for work groups to communicate and coordinate with each other during activities in which interdepartmental coordination was necessary to assure human performance [H.3(b)](Section 2RS02) Inspection Report# : 2012005 (pdf) Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary. Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Use of Waivers to Allow Workers to Exceed the Minimum Day Off Rule Page 15 of 16

1Q/2013 Inspection Findings - Grand Gulf 1 Green. The inspectors identified a non-cited violation of 10 CFR 26, Subpart I, Managing Fatigue, Subsection 207, Waivers and Exceptions, when the licensee inappropriately used waivers to allow workers to exceed the minimum day off rule. While reviewing condition reports, the inspectors noted the use of work hour waivers for a large number of staff. The circumstances for the use of waivers were the refueling outage lasting more than 60 days, contract expiration leading to 14 layoffs, and the loss of 4 workers via voluntary resignation. Due to these circumstances, work hours and fatigue of waivered individuals would have to be assessed daily. The assessment is required because the work hour limit of these individuals exceeded the minimum day off rule, therefore requiring daily monitoring until the end of the cycle. The waivered individuals averaged two days off per six-week period compared to the required three days off. Title 10 CFR 26.207 (a)(2) allows the granting of waivers only to address circumstances that could not have been reasonably controlled. The inspectors determined that the licensee was aware of the circumstances of an extended refueling outage and contract renewal deadline well in advance of the need to grant waivers, and a reasonable amount of time was available for the licensee to develop and execute contingency plans to negate the need to use waivers. Corrective actions included initiating assessments and waivers for exceeding minimum days off requirements for shift personnel for the six-week period ending May 27, 2012, and returning to the normal on-line work schedule in which adequate manpower is available to meet the requirements of the rule. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2012-7348. The finding is more than minor because it is associated with the access authorization attribute of the Security Cornerstone, and affected the cornerstone objective to provide assurance that the licensees security system and material control and accounting program use a defense in-depth approach and can protect against (1) the design basis threat of radiological sabotage from external and internal threats, and (2) the theft or loss of radiological materials. Using the Inspection Manual Chapter 0609, Appendix E, Baseline Security Significance Determination Process for Power Reactors, Figures 5 and 6, the finding was determined to have very low security significance because the calculated point total did not exceed the threshold value for a Green non-cited violation. The cumulative total for this finding was zero points, which was calculated by factoring the one impact area (vital areas) against Tier III Element 08.02.08, security force work hours, of the access authorization attribute, which resulted in a total of zero points within this attribute. The finding was determined to have a cross-cutting aspect in the area of human performance associated with the decision making component in that the licensee failed to use conservative assumptions in developing staff schedules for the duration of refueling outage 18 and for allowing an employment contract to expire that led to 14 individuals being laid off without realizing the impact these decisions would have on the licensees ability to meet the requirements of the rule [H.1(b)] (Section 1R20). Inspection Report# : 2012003 (pdf) Last modified : June 04, 2013 Page 16 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Grand Gulf 1 2Q/2013 Plant Inspection Findings Initiating Events Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: FIN Finding Reactor Scram Due to Ground Fault Green. The inspectors reviewed a self-revealing finding for the failure to ensure the current transformer structure, the neutral bus housing, and the associated mounting hardware were installed with adequate clearance to accommodate thermal expansion. This failure resulted in an automatic reactor scram on December 29, 2012, and a subsequent scram on January 4, 2013. Following the second scram on January 4, 2012, the licensee determined the cause of the scram was a trip of the phase A unit differential relay because of a ground fault on the A phase of the generator neutral current transformer, due to inadequate clearances. Immediate corrective actions included removing the damaged current transformer and modifying the neutral bus housing. The plant scrams were entered into the corrective action program as Condition Reports CR-GGN-2012-13290 and CR-GGN-2013-00083. The failure to install micarta plate bolts in accordance with manufacturer specifications and ensure that the current transformer structure, the neutral bus housing, and the associated mounting hardware had adequate clearance is a performance deficiency. This finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because it caused only a reactor trip and did not cause a loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to provide adequate work instructions [H.2(c)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedure for Removal of a Foreign Material Exclusion Plug Green. The inspectors reviewed a self-revealing non-cited violation of 10 CFR 50 Appendix B Criterion V, for the failure to provide adequate instructions to remove foreign material from the exhaust port of relief valve 1B21F047A. As a result, the valve failed to close at its reset setpoint following a reactor scram on December 29, 2012. The valve failed to close at its reset setpoint of 1013 psig and remained open until pressure fell to approximately 675 psig. The immediate corrective actions were to remove the foreign material exclusion plug from the exhaust port of valve 1B21-F047A and to ensure no plug was installed in any other safety relief valve. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-00100. The failure to provide adequate instructions to remove foreign material from the exhaust port of relief valve 1B21F047A is a performance deficiency. This finding is more than minor because it is associated with the Initiating Page 1 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because after a reasonable assessment of the degradation, the finding could not result in exceeding the reactor coolant leak rate for a small loss of coolant accident because the configuration of the safety relief valve was such that it would close at approximately 675 psig. Also the finding did not affect other systems used to mitigate a loss of coolant accident resulting in a total loss of their function. The finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee did not use a systematic process to make a safety-significant decision. [H.1 (a)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: FIN Finding Reactor Scram Due to Moisture in Isophase Bus Duct Green. The inspectors reviewed a self-revealing finding for the failure to identify a degraded isophase bus duct view port window, which allowed water to intrude into the duct and caused an automatic reactor scram on January 14, 2013. The licensee took corrective action to stop the water intrusion into the isophase bus duct and to electrically isolate the spare transformer from the energized transformers. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-00319. The failure to identify a degraded isophase bus duct view port window is a performance deficiency. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using NRC Inspection Manual Chapter 0609, , "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has a very low safety significance (Green) because it caused only a reactor trip and did not cause a loss of mitigating equipment relied on to transition the plant from the onset of a trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee did not use conservative assumptions in decision-making [H.1(b)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Revise the Scram Procedure After Temporary Modification Green. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the failure to revise the scram procedure after temporarily modifying the division-2 circuits that sense first-stage turbine pressure. Specifically, after a steam sensing line failed, the licensee had introduced a dummy signal into the subject circuits to comply with technical specifications; however, they failed to revise Procedure 05-1-02-I-1, Reactor Scram, Revision 117, to reflect this temporary modification. This resulted in additional scrams during scram recovery for the scrams on December 29, 2012, and January 4, 2013. Immediate corrective actions included modifying the scram procedure to require the operators to turn off the units that provide the dummy signal to the division-2 circuits that Page 2 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 sense first-stage turbine pressure following a reactor scram, allowing the operators to reset the full scram promptly. The licensee entered this issue into the corrective action program as Condition Report CR GGN-2013-001259. The failure to revise Procedure 05-1-02-I-1 following a temporary modification to the division-2 circuits that sense first-stage turbine pressure is a performance deficiency. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because it only caused a reactor trip and did not cause the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to ensure that procedures impacted by a temporary modification were properly revised to compensate for the installed modification [H.4(b)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Risk Significances and Develop Action Plans to Address Equipment Identified During Extent of Condition Review for a Post Scram Root Cause Analysis Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to follow procedure EN-LI-118, Root Cause Evaluation Process, Revision 18, in that they failed to evaluate the risk significances and develop action plans to address equipment identified during their extent-of-condition review for a post-scram root cause analysis. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11950. The immediate corrective actions included assigning corrective actions for operations personnel to properly evaluate the risk significance of the identified components and perform appropriate corrective actions to correct the degraded conditions. The licensees failure to properly determine risk significance and associated action plans to correct degraded equipment that could challenge safe plant operation is a performance deficiency. The performance deficiency is more than minor and is therefore a finding because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to take corrective actions to correct degraded equipment has the potential to lead to initiating events resulting in plant transients. Using NRC Inspection Manual Chapter 0609, , "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the apparent cause of this finding was that when operations management directed operators to identify the degraded equipment, they did not encourage those operators to comply with Procedure EN-LI-118. Therefore, the finding has a cross-cutting aspect in the human performance area, work practices component because the licensee did not define and effectively communicate expectations regarding procedural compliance. [H.4 (b)] (Section 4OA3). Inspection Report# : 2012005 (pdf) Page 3 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Mitigating Systems Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Seal Safety-related Manholes Green. The inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, involving the failure to ensure that manhole MH01 was properly sealed to prevent entry of flammable liquid. Specifically, on February 20, 2013, four manhole covers had between one to three loose bolts and evidence of water seepage. These vaults contain safety related cables for standby service water trains A and B. Immediate corrective actions included cleaning and tapping the bolt holes to ensure proper thread engagement, adding work instructions to the preventative maintenance procedure to clean the manhole bolt holes, and verifying that the other manholes containing safety-related cables did not have similar issues with loose bolts on the manhole covers. The licensee entered this issue in their corrective action program as Condition Report CR GGN-2013-01348. This finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of protection against external factors and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and required the use of Inspection Manual Chapter 0609, , Appendix F, Fire Protection Significance Determination Process. However, an NRC senior reactor analyst determined that the unique nature of this performance deficiency did not lend itself to analysis by the methods provided in Appendix F. Therefore, a Phase 3 analysis was performed. Based on a bounding analysis, the analyst determined that the change in core damage frequency was approximately 1.5E-7/yr. The result was low because of the relatively short periods of time that fuel was actually being transferred, the low probability of transfer system failures, and the low likelihood that a loss of normal service water initiator would occur following a fire in the subject manholes. The finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee did not provide adequate work packages [H.2(c)] (Section 1R06). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor for Ice on Standby Service Water Towers Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, for the licensees failure to monitor for ice accumulation on the standby service water cooling towers in accordance with station procedures. On January 17, 2013, the plant experienced a winter storm but operators did not implement Standby Service Water System Operating Instruction, 04-1-01-P41-1, Revision 137, Section 6.2, Cold Weather Operation, which directed the licensee to monitor the standby service water cooling tower for ice accumulation when weather conditions existed that could have resulted in icing of the cooling tower fill material and missile grating. The licensee entered this issue into their corrective action program as Condition Report CR-GGNS-2013-00426. The failure to monitor for ice accumulation in accordance with station procedures is a performance deficiency. The finding is more than minor because if left uncorrected, it could lead to a more significant safety concern. Specifically, the occurrence of ice accumulation on the standby service water cooling towers, if unmonitored, could cause damage to the fill material and/or the tower missile gratings, which would render the standby service water system inoperable. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Page 4 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency affecting the design or qualification of a mitigating system, structure or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allow outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. The finding has a cross-cutting aspect in the human performance area associated with the work control component because the licensee failed to appropriately plan work activities based on environmental conditions that may impact plant structures, systems and components [H.3(a)] (Section 1R13). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control for Setpoint Calculations Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure of the licensee to maintain design control, incorporate, verify, and check new instrument drift values, and translate the design basis requirements for multiple allowable values and trip setpoints described in the technical specifications into setpoint calculations. During the review of condition reports associated with an operability review of the licensees transition from an 18- to 24-month operating cycle in August 2012, inspectors identified that the licensee failed to maintain design control of multiple setpoint calculations. In response to NRC inspector questioning, a licensee review of the calculations identified that three of the 14 calculations reviewed contained calculated allowable values that differed from the values contained in the Technical Specifications associated with Level 8 Narrow Range, Reactor Scram on High SDVP Water Level, and HPCS & RCIC Pump Suction Transfer on High Suppression Pool Level. An assessment of the calculations also determined that one other calculation contained an error that was introduced during the replacement of the high-pressure turbine rotor in a recent refueling outage, which would require a license amendment request. The licensee entered this condition in their corrective action program as CR-GGN-2013-00371. The failure to maintain design control, incorporate, verify, and check new instrument drift values, and translate the design basis requirements into multiple allowable values and trip setpoints described in the technical specifications into facility setpoint calculations is a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective of ensuring the capability of the safety-related system to respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of the offsite power supply operability or functionality. This finding has a cross-cutting aspect in the area of human performance decision-making because the licensee did not use a systematic decision making process and did not obtain interdisciplinary input on a risk significant decision [H.1(a)] (Section 1R15). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Scaffold Affecting Fire Brigade Access Green. The inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, for Page 5 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 the failure to identify and correct a condition adverse to fire protection. Specifically, the licensee failed to ensure that fire brigade members had sufficient access through a scaffold built in the diesel generator building hallway into the division-1 diesel generator room. The immediate corrective actions included removing the scaffold in the diesel generator building hallway. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2013-01679. The failure to take prompt corrective action to ensure adequate access for fire brigade members through installed scaffolding in the diesel generator building hallway to the division-1 diesel generator room is a performance deficiency. The finding is more than because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the inability for fire brigade members to gain access to safety related equipment in timely manner could result in preventing prompt extinguishing of fires. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because the finding involved a risk-significant fire area that had an automatic fire suppression system. The inspectors determined the apparent cause of this finding was that the licensee did not implement the corrective action program with a low threshold for identifying scaffolding that could impede fire brigade member response during a fire. Therefore the finding had a cross-cutting aspect in the problem identification and resolution area associated with the corrective action program component because the licensee failed to identify conditions adverse to fire protection [P.1(a)] (Section 1R22). Inspection Report# : 2013002 (pdf) Significance: Feb 27, 2013 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Aligning Nitrogen Backup to Automatic Deressurization System

  • Green. The team identified a Green non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensees procedures for aligning portable nitrogen bottles to the Instrument Air system for backup operation of Automatic Depressurization System (ADS) valves do not include a step to direct the pressure regulator outlet isolation valves to be opened. If these valves are left closed, the nitrogen bottles will remain isolated from the Instrument Air system.

The failure to include a procedural step to open the nitrogen regulator outlet isolation valves when aligning nitrogen to the ADS valve instrument air lines is a performance deficiency. The performance deficiency is more than minor and is therefore a finding because it is associated with the procedure quality attribute of the mitigating systems cornerstone and affects the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This deficiency could have significantly affected the operators ability to perform the activity affecting quality, in this case, aligning nitrogen as a backup to ADS valve instrument air. Using Inspection Manual Chapter 0609, Attachment 4, Initial Screening and Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, a Phase 1 screening was performed and determined that the finding required a detailed risk evaluation because the finding would have resulted in a loss of system safety function had the procedure been called upon. The senior reactor analyst performed a detailed risk evaluation using the Grand Gulf Standardized Plant Analysis Risk model Version 8.22, and the SPAR-H human reliability analysis method. This method resulted in an incremental conditional core damage probability of 7.0 x 10-6. However, the analyst noted that, given the specific performance deficiency, this method provided a bounding analysis. Therefore, the finding was assessed using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process using Qualitative Criteria. The analyst noted that licensee calculations and Page 6 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 surveillance of the accumulators and associated check valves indicated that accumulator pressure would remain available for much longer than the 6 hours suggested in the model. Additionally, the failure of the 21 safety-relief valves under this condition would not occur simultaneously, but would be staggered as a result of the depressurization of individual accumulators. This would provide additional indication, cues, and time for operators to identify and correct the valve alignment error. Finally, the SPAR model does not consider the potential for recovery of the instrument air system. Based on this additional qualitative information, the analyst determined that the additional cues and time provided to the operators combined with the straight-forward diagnosis for this specific finding would reduce the overall risk of this performance deficiency by more than an order of magnitude. Therefore, using a bounding quantitative evaluation combined with qualitative factors, this finding was determined to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues by missing multiple opportunities to identify the procedural discrepancy when it was developing and validating the exam material for submission to the NRC [P.1(a)]. Inspection Report# : 2012301 (pdf) Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Adequate Procedure Instructions to Perform Preventive Maintenance Requiring The Periodic Replacement of the Control Relays in the GE Magne Blast Circuit Breakers Green. The inspectors reviewed a self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to complete preventive maintenance tasks on the high pressure core spray division III diesel generator output breaker in accordance with the corresponding preventive maintenance task template. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2012-07992. The immediate corrective actions included replacing the failed control relay and restoring operability to the division III diesel generator. The long term corrective actions included revising breaker refurbishment/replacement procedure with directions to replace the control relay and change the procedure frequency to every 10 years versus every 12 years. The inspectors determined that this performance deficiency was more than minor and is therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this failed control relay caused the subject breaker to fail to close during the division III diesel generator monthly surveillance on June 5, 2012. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," to determine that the issue affected the Mitigating System Cornerstone. Because the finding pertained only to a degraded condition while the plant was shutdown, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Checklist 8, Cold Shutdown or Refueling Operation - Time to Boil > 2 Hours: RCS Level < 23 Above Top of Flange, to determine that the finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add RCS inventory when needed; did not significantly degrade the licensees ability to recover decay heat removal if lost; and did not affect the safety/relief valves (Green). The inspectors determined that the cause of this finding was a latent issue that is not reflective of current performance, therefore no cross-cutting aspect was identified. (Section 1R20.b). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Page 7 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Gain Settings on APRM and LPRM Instruments in Accordance with Design Requirements Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish the gain settings used on the power range neutron monitoring system in accordance with design requirements. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2013-00177. The immediate corrective actions included adjusting gain settings for their average power range monitor (APRM) instruments to indicate actual core thermal power as determined by the heat balance. In additioin, the licensee revised their neutron monitoring procedure to set the initial gains for the average power range monitor to the maximum value to maintain conservative power indication during future startups. They also changed their local power range monitor replacement procedure to use the vendor specified initial gain setting of 3.692 prior to startup. The finding was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect gain settings caused a violation of technical specification 3.0.4 by rendering the APRM Neutron Flux High - Setdown scram function and the Neutron Flux - Upscale, Startup control rod block function inoperable prior to entry into Mode 2. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had very low safety significance (Green) because although the finding affected a single reactor protection system trip signal to initiate a reactor scram, it did not affect the function of other redundant trips or diverse methods of reactor shutdown, did not involve control manipulations that unintentionally added positive reactivity, and did not result in a mismanagement of reactivity by operators. Because the performance deficiency occurred in the past and is not reflective of current licensee performance, this finding was not assigned a cross-cutting aspect. (Section 4OA3). Inspection Report# : 2012005 (pdf) Significance: Sep 21, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Operability Determinations Green. The inspectors identified two examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure EN-OP-104, Operability Determinations. Specifically, for Condition Report CR-GGN-2012-09690, which documents an oil leak on the standby liquid control pump B, and for Condition Report CR-GGN-2012-09889, which documents degraded bolts on a flanged connection on standby service water B piping, the licensee failed to validate that operability evaluations completed for prior non-conforming conditions bounded the conditions documented in the new condition reports. As immediate corrective actions, the licensee re-performed the evaluations and established an adequate basis for operability for the conditions described in the two condition reports listed above. The licensee entered this issue into their corrective action program as CR-GGN-2012-09735 and CR-GGN-2012-10664. The finding was more than minor because if left uncorrected, not performing operability determinations in accordance with procedure could lead to a more significant safety concern. Specifically, if a condition renders a safety related system inoperable and because of this performance deficiency the licensee incorrectly determines that the system is operable, then this performance deficiency could result in a safety related system remaining inoperable for a long period of time. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because although it affected the design or qualification of a mitigating Page 8 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 system, the system maintained its operability. The finding had a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly evaluate for operability conditions adverse to quality [P.1(c)] (Section 1R15). Inspection Report# : 2012004 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of 4160 Vac Circuit Breakers for As-Found Tests Green. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to July 27, 2012, the licensees preventive maintenance Procedures 07-S-12-41, 07-S-12-42, and 07-S-12-61 failed to assure that the 4160 Vac circuit breakers would perform satisfactorily in service when the licensee performed maintenance prior to completing as-found tests to verify past operability of the circuit breakers. This finding has been entered into licensees corrective action program as Condition Reports CR-GGN-2012-09035 and CR- GGN-2012-9103. The team determined that failure to establish a test program which ensures that test and maintenance procedures associated with safety-related 4160 Vac circuit breakers would perform satisfactorily in service was a performance deficiency. This finding was more than minor because, if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to perform as-found tests prior to performing maintenance in preventive maintenance procedures was a significant programmatic deficiency which could cause unacceptable conditions to go undetected. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee failed to ensure that test and maintenance procedures were complete, accurate, and up-to-date to assure nuclear safety. [H.2(c)] (1R21.2.1) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish a Testing Program for Safety Related 125 Vdc Circuit Breakers Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, prior to July 27, 2012, the licensee failed to establish a test program for 125 Vdc safety related molded case circuit breakers incorporating the requirements of IEEE 308, to ensure the breakers would not degrade and would perform satisfactorily in service. The finding was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-09030 and CR-GGN-2012-09175. The team determined that the failure to establish a testing program incorporating the requirements of IEEE 308 was a performance deficiency. The finding was more than minor, because if left uncorrected, it would lead to a more Page 9 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 significant safety concern. Specifically, the failure to establish a testing program was a significant programmatic deficiency that would lead to missed opportunities to detect potential common cause failures from degradation of performance in more than one redundant safety division. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component; because the licensee failed to thoroughly evaluate problems such that resolutions address cause and extent of condition. Specifically, the licensee failed to thoroughly evaluate the extent of condition associated with previously identified NRC violation involving the failure to test 480 Vac molded case circuit breakers identified during the 2009 component design basis inspection. [P.1(c)] (1R21.2.2) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for a Change to Credible Passive Failures in the Standby Service Water System Severity Level IV. The team identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests and Experiments which states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if this activity would; result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report (as updated). Specifically, on August 18, 1987, the licensee implemented a change to the updated safety analysis report which limited credible passive failures in the standby service water system to pump and valve seal leakage without obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR GGN 2012 09267. The team determined that the licensees failure to receive prior NRC approval for changes in licensed activities regarding single passive failure criteria for the standby service water system was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding had the ability to impact the regulatory process. The performance deficiency was more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation. In accordance with the NRC Enforcement Manual, risk insights from the Inspection Manual Chapter 0609, Significance Determination Process, are used in determining the significance of 10 CFR 50.59 violations. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because, although the standby service water system could not provide 30 days of decay heat removal without operator action to provide makeup water to the system, it would have been able to complete its 24-hour risk significant mission time. Since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. The finding does not have a crosscutting aspect because the most significant contributor to the finding does not reflect current licensee performance. (1R21.2.3) Page 10 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Promptly Enter an NRC Violation Regarding the Standby Service Water System into the Corrective Action Program Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, on July 12, 2012, the NRC informed the licensee of a violation of 10 CFR 50.59 requirements, but the licensee failed to promptly identify this as an adverse condition and enter this condition into their corrective action program until July 19, 2012. The finding was entered into the licensees corrective action program as CR-GGN-2012-10075. The team determined that the licensees failure to promptly enter the NRC violation as condition adverse to quality into the corrective action program was a performance deficiency. This finding was more than minor because it adversely affected the design control attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to promptly document a violation of 10 CFR 50.59, which delayed an operability evaluation that ultimately determined that compensatory measures were required to ensure that the standby service water system could perform its specified safety function for its entire mission time. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because, although the standby service water system could not provide 30 days of decay heat removal without operator action to provide makeup water to the system, it would have been able to complete its 24-hour risk significant mission time. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated, and that actions are taken to address safety issues, in a timely manner, commensurate with their safety significance. Specifically, the licensee did not implement a corrective action program with a low threshold for identifying issues completely, accurately, and in a timely manner commensurate with their safety significance. [P.1(a)] (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Operability Determination Process Procedure Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Page 11 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Procedures, and Drawings which states, in part, that Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, from July 19, 2012, to July 29, 2012, the licensee failed correctly evaluate the operability of the standby service water system with a degraded or nonconforming condition and failed to document a sound basis for a reasonable expectation of operability of the standby service water system as required by Procedure EN-OP-104, Operability Determination Process. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2012-09356. The team determined that the failure to implement the requirements of the operability determination process procedure was a performance deficiency. The finding was more than minor because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the standby service water system was incapable of performing its specified safety function for the entire 30-day mission time without compensatory measures. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team determined that the finding represented a loss of system safety function in that the standby service water system could not meet its 30-day mission time to provide decay heat removal. Therefore, a Detailed Risk Evaluation was necessary. In accordance with Manual Chapter 0609, Appendix A, Section 6, Detailed Risk Evaluation, the senior reactor analyst evaluated the risk of the degraded condition that resulted from the finding. According to the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events, Section 4.1, Mission Time Modeling, in most events, 24 hours is sufficient time to bring numerous resources to bear on core cooling. In some events, the choice is conservative and the analysis results are overestimates. Additionally, the analyst determined that Section 4.2 on increasing mission time was not applicable to the subject finding because the decrease in standby service water system water inventory would be obvious and there would be days to respond with makeup sources. Therefore, the analyst determined that the finding was of very low safety significance (Green) because the standby service water system could would have been able to complete its 24-hour risk significant mission time although it could not provide 30 days of decay heat removal without operator action to provide makeup water to the system. This finding had a crosscutting aspect in the area of human performance, decision making component, because the licensee did not make decisions that demonstrated that nuclear safety was an overriding priority. Specifically, the licensee did not make safety significant decisions using a systematic process to ensure safety is maintained. [H.1(a)] (1R21.2.3) Inspection Report# : 2012008 (pdf) Significance: Sep 10, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Test and Inspection Requirements for 4160 Vac Circuit Breakers into Preventive Maintenance Procedures Green. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, which states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design document. Specifically, prior to July 27, 2012, the licensees safety-related 4160 Vac circuit breaker preventive maintenance Procedures 07-S-12-41, 07-S-12-42, and 07 S 12-61 failed to incorporate inspection and test requirements for minimum voltage tests, reduced voltage tests, and inspection of auxiliary switch relay contacts as established in the licensees circuit breaker maintenance program. This condition was entered into the licensees corrective action program as Condition Reports CR GGN 2012-08885 and CR-GGN-2012-09111. The team determined that the failure to incorporate required tests and inspections into preventive maintenance procedures for safety related 4160 Vac circuit breakers was a performance deficiency. This finding was more than Page 12 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 minor because, if left uncorrected, it would lead to a more significant safety concern. Specifically, the failure to incorporate the testing, cleaning, and inspection requirements into preventive maintenance procedures were a significant programmatic deficiency which could cause unacceptable conditions to go undetected. Using the Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of safety function. This finding had a crosscutting aspect in the area of problem identification and resolution, operating experience component, because the licensee failed to use operating experience information, including vendor recommendations and internally generated lessons learned, to support plant safety. Specifically, the licensee did not implement and institutionalize operating experience through changes to processes, procedures, equipment, and training programs. [P.2(b)] (1R21.2.4) Inspection Report# : 2012008 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: VIO Violation Inadequate Corrective Action for a Leak on the Division II Emergency Diesel Generator Lube Oil Sump Green. The team identified a Green cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a leak on the Division II emergency diesel generator lube oil sump. Despite identification of the leak in 2004, ineffective attempts to repair the leak and previous identification by the NRC in 2009, the licensee dispositioned the leak as accept as-is without a full understanding of the lube oil sump leak and potential consequences. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8880. The condition was discovered and documented by the licensee in 2004. This finding was initially determined by the NRC to be a minor violation in 2009. Paragraph F of Section 2.10 of the NRC Enforcement Manual states in part that where a licensee does not take corrective action for a minor violation, the matter should be considered more than minor and associated with a green inspection finding and dispositioned in a cited or noncited violation, as appropriate. This finding is now determined to be more than minor because if left uncorrected the failure to restore the lube oil sump for the Division II emergency diesel generator to design conditions would have the potential to lead to a more significant safety concern, specifically, the leak could worsen and potentially affect operability of the emergency diesel generator. Due to the licensees failure to restore compliance within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy. This finding affects the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate this problem such that the resolutions addressed the causes [P.1(c)]. (Section 4OA2.5d) Inspection Report# : 2011006 (pdf) Barrier Integrity Significance: Dec 31, 2012 Identified By: NRC Page 13 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Item Type: NCV NonCited Violation Failure to Make Timely Corrective Actions to Repair the Degraded Auxiliary Building Water Intrusion Barrier Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to correct a condition adverse to quality in a timely manner. Specifically, the licensee failed to correct multiple degraded conditions associated with the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-10314. Corrective actions included generating Work Order 318398 and delegating funds to repair the water intrusion barrier at the next available opportunity. The finding is more than minor because if left uncorrected, the condition of a degraded auxiliary building water intrusion barrier could lead to a more significant safety concern. Specifically, continued degradation of the water intrusion barrier could lead to the auxiliary building (secondary containment) being degraded such that the standby gas treatment system would not be able to achieve and maintain the design negative pressure of 1/4 inch water column within 120 seconds. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding had very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that the apparent cause of this finding was that the licensee had failed to classify the degraded water intrusion barrier as a condition adverse to quality that warranted correction in a timely manner. Therefore, the finding has a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly classify conditions adverse to quality [P.1(c)](Section 1R12). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Condition of the Auxiliary Building Water Intrusion Barrier Green. The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2), for the failure to evaluate the condition of the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11740. Corrective actions included initiating Condition Report CR-GGN-2012-12286, in which the licensee concluded the degraded water intrusion barrier had experienced a Maintenance Rule Functional Failure and required further evaluation to determine if the barrier should be classified in 10 CFR 50.65 (a) (1). The finding is more than minor because if left uncorrected, the failure to adequately evaluate the condition of the auxiliary building water intrusion barrier in accordance with the maintenance rule program could lead to a more significant safety concern. Specifically, continued inadequate evaluation of the water intrusion barrier could compromise the integrity of the secondary containment function of the auxiliary building. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that this finding is a latent issue; therefore no cross cutting aspect was assigned (Section 1R12). Inspection Report# : 2012005 (pdf) Page 14 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Adequately Plan and Control Work Activities to Maintain ALARA Green. The inspector reviewed a self-revealing finding of very low safety significance because during the refueling outage 18 extended power upgrade, the licensee did not adequately plan and control work activities for the design and replacement of the new fuel pool cooling heat exchangers. Specifically, outage personnel did not perform adequate pre-outage walkdowns, which resulted in significant unplanned collective exposure. Actual collective dose and hours for Radiation Work Permit 2012-1086, Fuel Pool Cooling & Cleanup Heat Exchanger Replacement, was 23.9 person-rem and 12,237 RWP-hours, respectively. This is compared to the initial planned estimate of 3.74 person-rem and 1,905 RWP-hours. This finding and procedural concern was entered into the corrective action program as Condition Reports CR-GGNS-2012-09011 and CR-GGNS-2012-12398. The failure to appropriately use ALARA planning and controls procedures to prevent unplanned and unintended collective doses was a performance deficiency. This performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone attribute of Program and Process in that the failure to adequately implement ALARA procedures caused the collective radiation dose for the job activity to exceed the planned dose by more than 50 percent. In addition, this type of issue is addressed in Example 6.j of IMC 0612, Appendix E, Examples of Minor Issues. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this performance deficiency to be a finding of very low safety significance because although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 240 person-rem. This finding has a cross-cutting aspect in the human performance area, work control component, because the licensee failed to evaluate the impact of work scope change on human performance and interdepartmental communication and coordination prior to commencing work activities. Specifically, there was inappropriate coordination and communication of work activities between work groups [H.3(b)](Section 2RS02). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Follow the Radiation Work Permit Requirements During Reactor Cavity High Water Operations Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1 for failure to comply with radiological exposure controls specified in Radiation Work Permit 2012-1402, Refuel Floor High Water Activities. Specifically, radiation exposure controls in the RWP required the licensee to verify that fuel pool cleanup (demineralizers) was in-service, and if dose rates increased by more than 0.2 millirem/hour, change the resins. During reactor cavity operations, both fuel pool demineralizer trains were inoperable at least 25 days. In addition, the dryer separator pool and reactor cavity were isolated from the fuel pool clean up system. Consequently, general area radiation levels on the reactor cavity floor increased from 0.4 millirem/hour to 6.0 millirem/hour. The actual collective dose and hours for the work activity was 8.24 person-rem and 9,000 RWP-hours, respectively. This is compared to the planned initial estimate of 4.60 person-rem and 6,987 RWP-hours. This Radiation Work Permint and procedure violation was documented in the licensees corrective action program as Condition Reports CR-GGNS-2012-04288 and CR-GGNS-2012-12401. The licensees failure to comply with the RWP to prevent unplanned and unintended collective doses was a Page 15 of 16

2Q/2013 Inspection Findings - Grand Gulf 1 performance deficiency. This performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone attribute of Program and Process in that the failure to adequately implement ALARA procedures caused the collective radiation dose for the job activity to exceed the planned dose by more than 50 percent. In addition, this type of issue is addressed in Example 6.i of IMC 0612, Appendix E, Examples of Minor Issues. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this performance deficiency to be a non-cited violation of very low safety significance because although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 240 person-rem. The violation involved a cross-cutting aspect in the human performance area, work control component, because the licensee did not appropriately coordinate work activities by incorporating actions to address the need for work groups to communicate and coordinate with each other during activities in which interdepartmental coordination was necessary to assure human performance [H.3(b)](Section 2RS02) Inspection Report# : 2012005 (pdf) Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary. Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : September 03, 2013 Page 16 of 16

3Q/2013 Inspection Findings - Grand Gulf 1 Grand Gulf 1 3Q/2013 Plant Inspection Findings Initiating Events Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: FIN Finding Reactor Scram Due to Ground Fault Green. The inspectors reviewed a self-revealing finding for the failure to ensure the current transformer structure, the neutral bus housing, and the associated mounting hardware were installed with adequate clearance to accommodate thermal expansion. This failure resulted in an automatic reactor scram on December 29, 2012, and a subsequent scram on January 4, 2013. Following the second scram on January 4, 2012, the licensee determined the cause of the scram was a trip of the phase A unit differential relay because of a ground fault on the A phase of the generator neutral current transformer, due to inadequate clearances. Immediate corrective actions included removing the damaged current transformer and modifying the neutral bus housing. The plant scrams were entered into the corrective action program as Condition Reports CR-GGN-2012-13290 and CR-GGN-2013-00083. The failure to install micarta plate bolts in accordance with manufacturer specifications and ensure that the current transformer structure, the neutral bus housing, and the associated mounting hardware had adequate clearance is a performance deficiency. This finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because it caused only a reactor trip and did not cause a loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to provide adequate work instructions [H.2(c)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedure for Removal of a Foreign Material Exclusion Plug Green. The inspectors reviewed a self-revealing non-cited violation of 10 CFR 50 Appendix B Criterion V, for the failure to provide adequate instructions to remove foreign material from the exhaust port of relief valve 1B21F047A. As a result, the valve failed to close at its reset setpoint following a reactor scram on December 29, 2012. The valve failed to close at its reset setpoint of 1013 psig and remained open until pressure fell to approximately 675 psig. The immediate corrective actions were to remove the foreign material exclusion plug from the exhaust port of valve 1B21-F047A and to ensure no plug was installed in any other safety relief valve. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-00100. The failure to provide adequate instructions to remove foreign material from the exhaust port of relief valve 1B21F047A is a performance deficiency. This finding is more than minor because it is associated with the Initiating Page 1 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because after a reasonable assessment of the degradation, the finding could not result in exceeding the reactor coolant leak rate for a small loss of coolant accident because the configuration of the safety relief valve was such that it would close at approximately 675 psig. Also the finding did not affect other systems used to mitigate a loss of coolant accident resulting in a total loss of their function. The finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee did not use a systematic process to make a safety-significant decision. [H.1 (a)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: FIN Finding Reactor Scram Due to Moisture in Isophase Bus Duct Green. The inspectors reviewed a self-revealing finding for the failure to identify a degraded isophase bus duct view port window, which allowed water to intrude into the duct and caused an automatic reactor scram on January 14, 2013. The licensee took corrective action to stop the water intrusion into the isophase bus duct and to electrically isolate the spare transformer from the energized transformers. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-00319. The failure to identify a degraded isophase bus duct view port window is a performance deficiency. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using NRC Inspection Manual Chapter 0609, , "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has a very low safety significance (Green) because it caused only a reactor trip and did not cause a loss of mitigating equipment relied on to transition the plant from the onset of a trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee did not use conservative assumptions in decision-making [H.1(b)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Revise the Scram Procedure After Temporary Modification Green. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the failure to revise the scram procedure after temporarily modifying the division-2 circuits that sense first-stage turbine pressure. Specifically, after a steam sensing line failed, the licensee had introduced a dummy signal into the subject circuits to comply with technical specifications; however, they failed to revise Procedure 05-1-02-I-1, Reactor Scram, Revision 117, to reflect this temporary modification. This resulted in additional scrams during scram recovery for the scrams on December 29, 2012, and January 4, 2013. Immediate corrective actions included modifying the scram procedure to require the operators to turn off the units that provide the dummy signal to the division-2 circuits that Page 2 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 sense first-stage turbine pressure following a reactor scram, allowing the operators to reset the full scram promptly. The licensee entered this issue into the corrective action program as Condition Report CR GGN-2013-001259. The failure to revise Procedure 05-1-02-I-1 following a temporary modification to the division-2 circuits that sense first-stage turbine pressure is a performance deficiency. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because it only caused a reactor trip and did not cause the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to ensure that procedures impacted by a temporary modification were properly revised to compensate for the installed modification [H.4(b)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Risk Significances and Develop Action Plans to Address Equipment Identified During Extent of Condition Review for a Post Scram Root Cause Analysis Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the licensees failure to follow procedure EN-LI-118, Root Cause Evaluation Process, Revision 18, in that they failed to evaluate the risk significances and develop action plans to address equipment identified during their extent-of-condition review for a post-scram root cause analysis. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11950. The immediate corrective actions included assigning corrective actions for operations personnel to properly evaluate the risk significance of the identified components and perform appropriate corrective actions to correct the degraded conditions. The licensees failure to properly determine risk significance and associated action plans to correct degraded equipment that could challenge safe plant operation is a performance deficiency. The performance deficiency is more than minor and is therefore a finding because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to take corrective actions to correct degraded equipment has the potential to lead to initiating events resulting in plant transients. Using NRC Inspection Manual Chapter 0609, , "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the apparent cause of this finding was that when operations management directed operators to identify the degraded equipment, they did not encourage those operators to comply with Procedure EN-LI-118. Therefore, the finding has a cross-cutting aspect in the human performance area, work practices component because the licensee did not define and effectively communicate expectations regarding procedural compliance. [H.4 (b)] (Section 4OA3). Inspection Report# : 2012005 (pdf) Page 3 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 Mitigating Systems Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Implement a Compensatory Fire Watch per Station Fire Protection Procedures Green. The inspectors identified a non-cited violation of Facility Operating License Condition 2.C(41) for the failure to properly implement a compensatory fire watch per the station fire protection program. Following an inadvertent release of carbon dioxide from the Cardox automatic fire suppression system into a division 2 safety related switchgear room located in the auxiliary building, the operators isolated the auxiliary building from the Cardox system to prevent any future inadvertent releases. The inspectors accompanied the fire watch patrol, which was required due to the isolation of the Cardox system to the auxiliary building, and they noted that during the patrol, none of the 10 rooms requiring a fire watch were checked. The inspectors brought this concern to the shift manager who confirmed that each room was required to be checked per the established fire watch criteria and that the fire watch patrol misunderstood the requirement. The licensee took immediate corrective action to direct the fire watch to check all the rooms to restore compliance with the fire watch requirements. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-04058. The failure to perform a fire watch in accordance with the fire protection program is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it is associated with the protection against the external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to perform the fire watch correctly adversely impacted the plants capability to detect and suppress a fire in a timely manner. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, , Table 3, the inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The inspectors determined that the finding had an adverse affect on the fixed fire protection systems element of fire watches posted as a compensatory measure for outages or degradations. The inspectors assigned a high degradation rating due to the automatic fire suppression system being tagged out of service. Because the system was degraded without compensatory actions for less than three days, the inspectors used a duration factor of 0.01. The inspectors used 2E-2 for a generic fire frequency area for a switchgear room. The resulting change in core damage frequency was 2E-4, which was greater than the high degradation Phase 1 Quantitative Screening Criteria of 1E-6. Therefore, a senior reactor analyst performed a detailed risk evaluation. The analyst performed a bounding analysis of the performance deficiency (See Table 1R05-1). For each of the 10 affected fire areas, the analyst determined the probability of a fire occurring by multiplying the fire ignition frequency from the licensees fire hazards analysis by the 9.2 hours that the performance deficiency impacted the plant. Because each fire area had a functional fire detection system throughout the exposure period, the analyst determined the non-detection probability by multiplying the fire probability by the generic failure probability for a detection system. The analyst made the bounding assumption that all fires postulated to initiate that were not detected would proceed to core damage. The sum of all the non-detection probabilities was 9.1 x 10-7 (See Table 1R05-1). Therefore, the bounding analysis indicates that this finding is of very low safety significance (Green). The inspectors determined the apparent cause of this finding was that the security officers performing the fire watch patrols did not understand the requirement to visually check the affected rooms. Therefore, the finding has a cross-cutting aspect in the human performance area associated with the work practices component because the licensee did not communicate human error prevention techniques such as pre-job briefings and proper documentation of activities commensurate with the risk of the assigned task [H.4(a)] (Section 1R05.1.b). Inspection Report# : 2013003 (pdf) Page 4 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Nonconforming Condition with the Standby Diesel Generator Inlet Plenum Turning Vanes Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. Specifically, from November 3, 1988, until March 6, 2010, actions to correct known design deficiencies on the left and right banks intercooler inlet plenums of both the division 1 and 2 standby diesel generators were not fully implemented. The design deficiency, identified by the vendor, could result in intercooler tube failure and jacket water leakage. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2013-02631. The failure to correct a nonconforming condition in the division 1 and 2 standby diesel generators inlet plenums is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it adversely affected the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage), and if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to implement corrective actions to resolve a known design deficiency of the intercooler inlet plenums could have resulted in either the division 1 or 2 standby diesel generator failing to perform its safety function. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green), because the finding was a design deficiency affecting a mitigating systems structure, system, or component that did not lose operability or functionality. The finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance (Section 1R17.1.b.2). Inspection Report# : 2013003 (pdf) Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Nonconforming Condition in the Train B Starting Circuit Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. Specifically, from November 20, 1998, until November 7, 2012, actions to correct a known nonconforming condition involving the low pressure interlock of the train B starting circuit on both the division 1 and 2 standby diesel generators had not been implemented. If the train A starting circuit were to fail and starting air pressure were to fall below 120 psig, the diesel generator would have all automatic shutdown permissives active, which are supposed to be bypassed during a loss-of-coolant-accident signal. This was considered a single point vulnerability for the train B starting circuit. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2013-02524. The failure to correct a nonconforming condition in the division 1 and 2 standby diesel generators train B starting circuits is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it affected the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage), and if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to implement corrective actions to resolve a known nonconforming condition of the low pressure interlock of the train B starting circuit could have resulted in either the division 1 or 2 standby diesel generator failing to perform its safety function. Using Inspection Manual Page 5 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green), because the finding was a design deficiency affecting a mitigating systems structure, system, or component that did not lose operability or functionality. The finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance (Section 1R17.1.b.3). Inspection Report# : 2013003 (pdf) Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify the Residual Heat Removal B System was Full of Water Within its Specified Frequency Green. The inspectors identified a non-cited violation of Technical Specification Surveillance Requirement SR 3.5.1.1 for the failure to verify the residual heat removal B system was full of water within its specified frequency. The inspectors reviewed a low pressure core injection subsystem B monthly functional test that was performed on April 10, 2013, per Procedure 06-OP-1E12-M-0002, LPCI/RHR Subsystem B Monthly Functional Test, Revision 113. The inspectors identified that the licensee failed to perform ultra sonic testing on the pipe prior to and after venting of the pipe directly upstream of the residual heat removal heat exchanger B outboard vent valve (1E12F074B). By not performing the ultra sonic testing, the licensee did not verify the residual heat removal system was full of water as required by Surveillance Requirement 3.5.1.1. Immediate corrective actions included performing the ultra sonic tests, which verified the system was full of water and satisfied the surveillance requirement. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2013-02847. The failure to verify the residual heat removal B system was full of water as required by Technical Specification Surveillance Requirement SR 3.5.1.1 is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstones objective of ensuring the availability, reliability and capability of systems that respond to prevent undesirable consequences. Specifically, the failure to perform the required ultra sonic testing resulted in Technical Specification Surveillance Requirement SR 3.5.1.1 not being met. Therefore, the licensee could not ensure the system would perform properly by injecting its full capacity into the reactor coolant system upon demand. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency affecting the design or qualification of a mitigating system, structure, or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allowed outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. Through interviews with operations personnel, the inspectors determined the apparent cause of the finding was that management failed to ensure the ultra sonic test was performed. Therefore, the finding had a cross-cutting aspect in the human performance area associated with the work practices component because the licensee failed to ensure supervisory and management oversight of work activities [H.4(c)] (Section 1R22.b). Inspection Report# : 2013003 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Seal Safety-related Manholes Green. The inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, involving the failure to ensure that manhole MH01 was properly sealed to prevent entry of flammable liquid. Specifically, on February 20, 2013, four manhole covers had between one to three loose bolts and evidence of water Page 6 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 seepage. These vaults contain safety related cables for standby service water trains A and B. Immediate corrective actions included cleaning and tapping the bolt holes to ensure proper thread engagement, adding work instructions to the preventative maintenance procedure to clean the manhole bolt holes, and verifying that the other manholes containing safety-related cables did not have similar issues with loose bolts on the manhole covers. The licensee entered this issue in their corrective action program as Condition Report CR GGN-2013-01348. This finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of protection against external factors and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and required the use of Inspection Manual Chapter 0609, , Appendix F, Fire Protection Significance Determination Process. However, an NRC senior reactor analyst determined that the unique nature of this performance deficiency did not lend itself to analysis by the methods provided in Appendix F. Therefore, a Phase 3 analysis was performed. Based on a bounding analysis, the analyst determined that the change in core damage frequency was approximately 1.5E-7/yr. The result was low because of the relatively short periods of time that fuel was actually being transferred, the low probability of transfer system failures, and the low likelihood that a loss of normal service water initiator would occur following a fire in the subject manholes. The finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee did not provide adequate work packages [H.2(c)] (Section 1R06). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor for Ice on Standby Service Water Towers Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, for the licensees failure to monitor for ice accumulation on the standby service water cooling towers in accordance with station procedures. On January 17, 2013, the plant experienced a winter storm but operators did not implement Standby Service Water System Operating Instruction, 04-1-01-P41-1, Revision 137, Section 6.2, Cold Weather Operation, which directed the licensee to monitor the standby service water cooling tower for ice accumulation when weather conditions existed that could have resulted in icing of the cooling tower fill material and missile grating. The licensee entered this issue into their corrective action program as Condition Report CR-GGNS-2013-00426. The failure to monitor for ice accumulation in accordance with station procedures is a performance deficiency. The finding is more than minor because if left uncorrected, it could lead to a more significant safety concern. Specifically, the occurrence of ice accumulation on the standby service water cooling towers, if unmonitored, could cause damage to the fill material and/or the tower missile gratings, which would render the standby service water system inoperable. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency affecting the design or qualification of a mitigating system, structure or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allow outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. The finding has a cross-cutting aspect in the human performance area associated with the work control component because the licensee failed to appropriately plan work activities based on environmental conditions that may impact plant structures, systems and components [H.3(a)] (Section 1R13). Inspection Report# : 2013002 (pdf) Page 7 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control for Setpoint Calculations Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure of the licensee to maintain design control, incorporate, verify, and check new instrument drift values, and translate the design basis requirements for multiple allowable values and trip setpoints described in the technical specifications into setpoint calculations. During the review of condition reports associated with an operability review of the licensees transition from an 18- to 24-month operating cycle in August 2012, inspectors identified that the licensee failed to maintain design control of multiple setpoint calculations. In response to NRC inspector questioning, a licensee review of the calculations identified that three of the 14 calculations reviewed contained calculated allowable values that differed from the values contained in the Technical Specifications associated with Level 8 Narrow Range, Reactor Scram on High SDVP Water Level, and HPCS & RCIC Pump Suction Transfer on High Suppression Pool Level. An assessment of the calculations also determined that one other calculation contained an error that was introduced during the replacement of the high-pressure turbine rotor in a recent refueling outage, which would require a license amendment request. The licensee entered this condition in their corrective action program as CR-GGN-2013-00371. The failure to maintain design control, incorporate, verify, and check new instrument drift values, and translate the design basis requirements into multiple allowable values and trip setpoints described in the technical specifications into facility setpoint calculations is a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective of ensuring the capability of the safety-related system to respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of the offsite power supply operability or functionality. This finding has a cross-cutting aspect in the area of human performance decision-making because the licensee did not use a systematic decision making process and did not obtain interdisciplinary input on a risk significant decision [H.1(a)] (Section 1R15). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Scaffold Affecting Fire Brigade Access Green. The inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, for the failure to identify and correct a condition adverse to fire protection. Specifically, the licensee failed to ensure that fire brigade members had sufficient access through a scaffold built in the diesel generator building hallway into the division-1 diesel generator room. The immediate corrective actions included removing the scaffold in the diesel generator building hallway. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2013-01679. The failure to take prompt corrective action to ensure adequate access for fire brigade members through installed scaffolding in the diesel generator building hallway to the division-1 diesel generator room is a performance deficiency. The finding is more than because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the inability for fire brigade members to gain access to safety related equipment in timely manner could result in preventing prompt extinguishing of fires. Using NRC Inspection Manual Page 8 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because the finding involved a risk-significant fire area that had an automatic fire suppression system. The inspectors determined the apparent cause of this finding was that the licensee did not implement the corrective action program with a low threshold for identifying scaffolding that could impede fire brigade member response during a fire. Therefore the finding had a cross-cutting aspect in the problem identification and resolution area associated with the corrective action program component because the licensee failed to identify conditions adverse to fire protection [P.1(a)] (Section 1R22). Inspection Report# : 2013002 (pdf) Significance: Feb 27, 2013 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Aligning Nitrogen Backup to Automatic Deressurization System

  • Green. The team identified a Green non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensees procedures for aligning portable nitrogen bottles to the Instrument Air system for backup operation of Automatic Depressurization System (ADS) valves do not include a step to direct the pressure regulator outlet isolation valves to be opened. If these valves are left closed, the nitrogen bottles will remain isolated from the Instrument Air system.

The failure to include a procedural step to open the nitrogen regulator outlet isolation valves when aligning nitrogen to the ADS valve instrument air lines is a performance deficiency. The performance deficiency is more than minor and is therefore a finding because it is associated with the procedure quality attribute of the mitigating systems cornerstone and affects the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This deficiency could have significantly affected the operators ability to perform the activity affecting quality, in this case, aligning nitrogen as a backup to ADS valve instrument air. Using Inspection Manual Chapter 0609, Attachment 4, Initial Screening and Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, a Phase 1 screening was performed and determined that the finding required a detailed risk evaluation because the finding would have resulted in a loss of system safety function had the procedure been called upon. The senior reactor analyst performed a detailed risk evaluation using the Grand Gulf Standardized Plant Analysis Risk model Version 8.22, and the SPAR-H human reliability analysis method. This method resulted in an incremental conditional core damage probability of 7.0 x 10-6. However, the analyst noted that, given the specific performance deficiency, this method provided a bounding analysis. Therefore, the finding was assessed using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process using Qualitative Criteria. The analyst noted that licensee calculations and surveillance of the accumulators and associated check valves indicated that accumulator pressure would remain available for much longer than the 6 hours suggested in the model. Additionally, the failure of the 21 safety-relief valves under this condition would not occur simultaneously, but would be staggered as a result of the depressurization of individual accumulators. This would provide additional indication, cues, and time for operators to identify and correct the valve alignment error. Finally, the SPAR model does not consider the potential for recovery of the instrument air system. Based on this additional qualitative information, the analyst determined that the additional cues and time provided to the operators combined with the straight-forward diagnosis for this specific finding would reduce the overall risk of this performance deficiency by more than an order of magnitude. Therefore, using a bounding quantitative evaluation combined with qualitative factors, this finding was determined to be of very low safety significance (Green). Page 9 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues by missing multiple opportunities to identify the procedural discrepancy when it was developing and validating the exam material for submission to the NRC [P.1(a)]. Inspection Report# : 2012301 (pdf) Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Implement Adequate Procedure Instructions to Perform Preventive Maintenance Requiring The Periodic Replacement of the Control Relays in the GE Magne Blast Circuit Breakers Green. The inspectors reviewed a self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to complete preventive maintenance tasks on the high pressure core spray division III diesel generator output breaker in accordance with the corresponding preventive maintenance task template. The licensee entered this issue in their corrective action program as Condition Report CR-GGN-2012-07992. The immediate corrective actions included replacing the failed control relay and restoring operability to the division III diesel generator. The long term corrective actions included revising breaker refurbishment/replacement procedure with directions to replace the control relay and change the procedure frequency to every 10 years versus every 12 years. The inspectors determined that this performance deficiency was more than minor and is therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this failed control relay caused the subject breaker to fail to close during the division III diesel generator monthly surveillance on June 5, 2012. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," to determine that the issue affected the Mitigating System Cornerstone. Because the finding pertained only to a degraded condition while the plant was shutdown, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Checklist 8, Cold Shutdown or Refueling Operation - Time to Boil > 2 Hours: RCS Level < 23 Above Top of Flange, to determine that the finding was of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add RCS inventory when needed; did not significantly degrade the licensees ability to recover decay heat removal if lost; and did not affect the safety/relief valves (Green). The inspectors determined that the cause of this finding was a latent issue that is not reflective of current performance, therefore no cross-cutting aspect was identified. (Section 1R20.b). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Establish Gain Settings on APRM and LPRM Instruments in Accordance with Design Requirements Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish the gain settings used on the power range neutron monitoring system in accordance with design requirements. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2013-00177. The immediate corrective actions included adjusting gain settings for their average power range monitor (APRM) instruments to indicate actual core thermal power as determined by the heat balance. In additioin, the licensee revised their neutron monitoring procedure to set the initial gains for the average power range monitor to the maximum value to maintain conservative power indication during future startups. They also changed their local power range monitor replacement procedure to use the vendor specified initial gain setting of 3.692 prior to startup. Page 10 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 The finding was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the incorrect gain settings caused a violation of technical specification 3.0.4 by rendering the APRM Neutron Flux High - Setdown scram function and the Neutron Flux - Upscale, Startup control rod block function inoperable prior to entry into Mode 2. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had very low safety significance (Green) because although the finding affected a single reactor protection system trip signal to initiate a reactor scram, it did not affect the function of other redundant trips or diverse methods of reactor shutdown, did not involve control manipulations that unintentionally added positive reactivity, and did not result in a mismanagement of reactivity by operators. Because the performance deficiency occurred in the past and is not reflective of current licensee performance, this finding was not assigned a cross-cutting aspect. (Section 4OA3). Inspection Report# : 2012005 (pdf) Significance: Oct 21, 2011 Identified By: NRC Item Type: VIO Violation Inadequate Corrective Action for a Leak on the Division II Emergency Diesel Generator Lube Oil Sump Green. The team identified a Green cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a leak on the Division II emergency diesel generator lube oil sump. Despite identification of the leak in 2004, ineffective attempts to repair the leak and previous identification by the NRC in 2009, the licensee dispositioned the leak as accept as-is without a full understanding of the lube oil sump leak and potential consequences. The licensee entered this condition into their corrective action program as condition report CR-GGN-2011-8880. The condition was discovered and documented by the licensee in 2004. This finding was initially determined by the NRC to be a minor violation in 2009. Paragraph F of Section 2.10 of the NRC Enforcement Manual states in part that where a licensee does not take corrective action for a minor violation, the matter should be considered more than minor and associated with a green inspection finding and dispositioned in a cited or noncited violation, as appropriate. This finding is now determined to be more than minor because if left uncorrected the failure to restore the lube oil sump for the Division II emergency diesel generator to design conditions would have the potential to lead to a more significant safety concern, specifically, the leak could worsen and potentially affect operability of the emergency diesel generator. Due to the licensees failure to restore compliance within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy. This finding affects the mitigating systems cornerstone. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, this finding was of very low safety significance because it did not create a loss of system safety function of a single train for greater than the technical specification allowed outage times, and did not affect seismic, flooding, or severe weather initiating events. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate this problem such that the resolutions addressed the causes [P.1(c)]. (Section 4OA2.5d) Inspection Report# : 2011006 (pdf) Barrier Integrity Page 11 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Make Timely Corrective Actions to Repair the Degraded Auxiliary Building Water Intrusion Barrier Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to correct a condition adverse to quality in a timely manner. Specifically, the licensee failed to correct multiple degraded conditions associated with the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-10314. Corrective actions included generating Work Order 318398 and delegating funds to repair the water intrusion barrier at the next available opportunity. The finding is more than minor because if left uncorrected, the condition of a degraded auxiliary building water intrusion barrier could lead to a more significant safety concern. Specifically, continued degradation of the water intrusion barrier could lead to the auxiliary building (secondary containment) being degraded such that the standby gas treatment system would not be able to achieve and maintain the design negative pressure of 1/4 inch water column within 120 seconds. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding had very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that the apparent cause of this finding was that the licensee had failed to classify the degraded water intrusion barrier as a condition adverse to quality that warranted correction in a timely manner. Therefore, the finding has a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly classify conditions adverse to quality [P.1(c)](Section 1R12). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Monitor the Condition of the Auxiliary Building Water Intrusion Barrier Green. The inspectors identified a non-cited violation of 10 CFR 50.65(a)(2), for the failure to evaluate the condition of the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11740. Corrective actions included initiating Condition Report CR-GGN-2012-12286, in which the licensee concluded the degraded water intrusion barrier had experienced a Maintenance Rule Functional Failure and required further evaluation to determine if the barrier should be classified in 10 CFR 50.65 (a) (1). The finding is more than minor because if left uncorrected, the failure to adequately evaluate the condition of the auxiliary building water intrusion barrier in accordance with the maintenance rule program could lead to a more significant safety concern. Specifically, continued inadequate evaluation of the water intrusion barrier could compromise the integrity of the secondary containment function of the auxiliary building. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that this finding is a latent issue; therefore no cross cutting aspect was assigned (Section 1R12). Inspection Report# : 2012005 (pdf) Page 12 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding Failure to Adequately Plan and Control Work Activities to Maintain ALARA Green. The inspector reviewed a self-revealing finding of very low safety significance because during the refueling outage 18 extended power upgrade, the licensee did not adequately plan and control work activities for the design and replacement of the new fuel pool cooling heat exchangers. Specifically, outage personnel did not perform adequate pre-outage walkdowns, which resulted in significant unplanned collective exposure. Actual collective dose and hours for Radiation Work Permit 2012-1086, Fuel Pool Cooling & Cleanup Heat Exchanger Replacement, was 23.9 person-rem and 12,237 RWP-hours, respectively. This is compared to the initial planned estimate of 3.74 person-rem and 1,905 RWP-hours. This finding and procedural concern was entered into the corrective action program as Condition Reports CR-GGNS-2012-09011 and CR-GGNS-2012-12398. The failure to appropriately use ALARA planning and controls procedures to prevent unplanned and unintended collective doses was a performance deficiency. This performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone attribute of Program and Process in that the failure to adequately implement ALARA procedures caused the collective radiation dose for the job activity to exceed the planned dose by more than 50 percent. In addition, this type of issue is addressed in Example 6.j of IMC 0612, Appendix E, Examples of Minor Issues. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this performance deficiency to be a finding of very low safety significance because although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 240 person-rem. This finding has a cross-cutting aspect in the human performance area, work control component, because the licensee failed to evaluate the impact of work scope change on human performance and interdepartmental communication and coordination prior to commencing work activities. Specifically, there was inappropriate coordination and communication of work activities between work groups [H.3(b)](Section 2RS02). Inspection Report# : 2012005 (pdf) Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Follow the Radiation Work Permit Requirements During Reactor Cavity High Water Operations Green. The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1 for failure to comply with radiological exposure controls specified in Radiation Work Permit 2012-1402, Refuel Floor High Water Activities. Specifically, radiation exposure controls in the RWP required the licensee to verify that fuel pool cleanup (demineralizers) was in-service, and if dose rates increased by more than 0.2 millirem/hour, change the resins. During reactor cavity operations, both fuel pool demineralizer trains were inoperable at least 25 days. In addition, the dryer separator pool and reactor cavity were isolated from the fuel pool clean up system. Consequently, general area radiation levels on the reactor cavity floor increased from 0.4 millirem/hour to 6.0 millirem/hour. The actual collective dose and hours for the work activity was 8.24 person-rem and 9,000 RWP-hours, respectively. This is compared to the planned initial estimate of 4.60 person-rem and 6,987 RWP-hours. This Radiation Work Permint and procedure violation was documented in the licensees corrective action program as Condition Reports CR-GGNS-2012-04288 Page 13 of 14

3Q/2013 Inspection Findings - Grand Gulf 1 and CR-GGNS-2012-12401. The licensees failure to comply with the RWP to prevent unplanned and unintended collective doses was a performance deficiency. This performance deficiency was more than minor because it affected the Occupational Radiation Safety Cornerstone attribute of Program and Process in that the failure to adequately implement ALARA procedures caused the collective radiation dose for the job activity to exceed the planned dose by more than 50 percent. In addition, this type of issue is addressed in Example 6.i of IMC 0612, Appendix E, Examples of Minor Issues. Using the Occupational Radiation Safety Significance Determination Process, the inspector determined this performance deficiency to be a non-cited violation of very low safety significance because although it involved ALARA planning and controls, the licensees latest rolling three-year average does not exceed 240 person-rem. The violation involved a cross-cutting aspect in the human performance area, work control component, because the licensee did not appropriately coordinate work activities by incorporating actions to address the need for work groups to communicate and coordinate with each other during activities in which interdepartmental coordination was necessary to assure human performance [H.3(b)](Section 2RS02) Inspection Report# : 2012005 (pdf) Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary. Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed. Miscellaneous Last modified : December 03, 2013 Page 14 of 14

4Q/2013 Inspection Findings - Grand Gulf 1 Grand Gulf 1 4Q/2013 Plant Inspection Findings Initiating Events Significance: Sep 30, 2013 Identified By: NRC Item Type: FIN Finding Failure to Follow Alarm Response Steps to Restore the TSE Following Maintenance Green. The inspectors reviewed a Green self-revealing finding for the failure to follow Procedure 04-1-02-1H13-P680-9A, TSE INFL OFF, Revision 36; in that operations personnel did not verify steps were followed per this alarm response procedure prior to returning the turbine thermal stress evaluator (TSE) to service following maintenance activities. The failure to follow alarm response procedure then resulted in an automatic reactor scram on July 30, 2013. Site personnel determined that the scram was caused by high reactor pressure resulting from the turbine unloading beyond the capability of the bypass valves after restoring the TSE to service following maintenance. On July 26, 2013, the control room received an alarm "TSE-STU CAB FAIL." The licensee failed to determine the correct cause of the alarm due to inadequate troubleshooting. Therefore, when the maintenance was completed and the TSE was returned to service, the turbine started to unload resulting in a reactor scram due to reactor vessel high pressure. The immediate corrective actions included determining the cause of the scram and taking actions to restore equipment prior to plant startup. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2013-04943. The failure to follow alarm response steps to restore the TSE following maintenance is a performance deficiency. Specifically, Procedure 04-1-02-1H13-P680-9A, TSE INFL OFF, Revision 36, step 4.1 requires operational personnel to ensure that the TSE is functioning correctly following maintenance prior to restoring to service. The performance deficiency is more than minor, and therefore a finding, because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that the issue has a very low safety significance (Green) because it only caused a reactor trip and did not cause a loss of mitigating equipment relied on to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined that the apparent cause of the finding was that the licensee did not troubleshoot to validate the cause for alarm TSE STU Cab Failure in accordance with station troubleshooting procedures. Therefore, the finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee did not use the troubleshooting process effectively [H.4(b)] (Section 4OA3). Inspection Report# : 2013004 (pdf) Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: FIN Finding Reactor Scram Due to Ground Fault Page 1 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Green. The inspectors reviewed a self-revealing finding for the failure to ensure the current transformer structure, the neutral bus housing, and the associated mounting hardware were installed with adequate clearance to accommodate thermal expansion. This failure resulted in an automatic reactor scram on December 29, 2012, and a subsequent scram on January 4, 2013. Following the second scram on January 4, 2012, the licensee determined the cause of the scram was a trip of the phase A unit differential relay because of a ground fault on the A phase of the generator neutral current transformer, due to inadequate clearances. Immediate corrective actions included removing the damaged current transformer and modifying the neutral bus housing. The plant scrams were entered into the corrective action program as Condition Reports CR-GGN-2012-13290 and CR-GGN-2013-00083. The failure to install micarta plate bolts in accordance with manufacturer specifications and ensure that the current transformer structure, the neutral bus housing, and the associated mounting hardware had adequate clearance is a performance deficiency. This finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because it caused only a reactor trip and did not cause a loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to provide adequate work instructions [H.2(c)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Procedure for Removal of a Foreign Material Exclusion Plug Green. The inspectors reviewed a self-revealing non-cited violation of 10 CFR 50 Appendix B Criterion V, for the failure to provide adequate instructions to remove foreign material from the exhaust port of relief valve 1B21F047A. As a result, the valve failed to close at its reset setpoint following a reactor scram on December 29, 2012. The valve failed to close at its reset setpoint of 1013 psig and remained open until pressure fell to approximately 675 psig. The immediate corrective actions were to remove the foreign material exclusion plug from the exhaust port of valve 1B21-F047A and to ensure no plug was installed in any other safety relief valve. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-00100. The failure to provide adequate instructions to remove foreign material from the exhaust port of relief valve 1B21F047A is a performance deficiency. This finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because after a reasonable assessment of the degradation, the finding could not result in exceeding the reactor coolant leak rate for a small loss of coolant accident because the configuration of the safety relief valve was such that it would close at approximately 675 psig. Also the finding did not affect other systems used to mitigate a loss of coolant accident resulting in a total loss of their function. The finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee did not use a systematic process to make a safety-significant decision. [H.1 (a)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Page 2 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Significance: Mar 31, 2013 Identified By: Self-Revealing Item Type: FIN Finding Reactor Scram Due to Moisture in Isophase Bus Duct Green. The inspectors reviewed a self-revealing finding for the failure to identify a degraded isophase bus duct view port window, which allowed water to intrude into the duct and caused an automatic reactor scram on January 14, 2013. The licensee took corrective action to stop the water intrusion into the isophase bus duct and to electrically isolate the spare transformer from the energized transformers. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-00319. The failure to identify a degraded isophase bus duct view port window is a performance deficiency. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using NRC Inspection Manual Chapter 0609, , "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has a very low safety significance (Green) because it caused only a reactor trip and did not cause a loss of mitigating equipment relied on to transition the plant from the onset of a trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee did not use conservative assumptions in decision-making [H.1(b)] (Section 4OA3). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Revise the Scram Procedure After Temporary Modification Green. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the failure to revise the scram procedure after temporarily modifying the division-2 circuits that sense first-stage turbine pressure. Specifically, after a steam sensing line failed, the licensee had introduced a dummy signal into the subject circuits to comply with technical specifications; however, they failed to revise Procedure 05-1-02-I-1, Reactor Scram, Revision 117, to reflect this temporary modification. This resulted in additional scrams during scram recovery for the scrams on December 29, 2012, and January 4, 2013. Immediate corrective actions included modifying the scram procedure to require the operators to turn off the units that provide the dummy signal to the division-2 circuits that sense first-stage turbine pressure following a reactor scram, allowing the operators to reset the full scram promptly. The licensee entered this issue into the corrective action program as Condition Report CR GGN-2013-001259. The failure to revise Procedure 05-1-02-I-1 following a temporary modification to the division-2 circuits that sense first-stage turbine pressure is a performance deficiency. The finding is more than minor because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because it only caused a reactor trip and did not cause the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to ensure that procedures impacted by a temporary modification were properly revised to compensate for the installed modification [H.4(b)] (Section 4OA3). Page 3 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Inspection Report# : 2013002 (pdf) Mitigating Systems Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Operability Determination Green. The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure EN-OP-104, Operability Determinations. Specifically, the inspectors identified that the licensee failed to establish an adequate basis for operability when a degraded or nonconforming condition had been identified. On August 30, 2013, Condition Report CR-GGN-2013-05604 was initiated to document a step change in the standby service water (SSW) siphon line K factor, which is a measure of flow through the siphon line. The K factor could have increased due to air entrapment in the siphon line that resulted from using air to mix the basin water following chemical treatments. The inspectors challenged the validity of the evaluation because the second step change in K factor, from 48 to 64, represented new information that had not been evaluated in the previous condition report. As an immediate corrective action, the licensee re-performed the operability determination and provided an adequate basis of operability by evaluating the system with the additional K factor data. Furthermore, the licensee verified the siphon line did not have any obstructions by observing the SSW basin levels equalize as water flowed through the siphon line. The licensee entered this issue into the corrective action process under Condition Report CR-GGN-2013-05687. The failure to perform an operability determination in accordance with procedure was a performance deficiency. The performance deficiency was more than minor, and is therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the reliability, availability and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. The inspectors determined that the apparent cause of this finding was that the licensee had identified and used previously completed operability evaluations without verifying that the previously completed evaluations were fully applicable to the identified conditions. Therefore, the finding had a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly evaluate for operability conditions adverse to quality [P.1(c)] (Section 1R15). Inspection Report# : 2013004 (pdf) Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Review Temporary Modifications by Operations Personnel During Turnover Green. The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Page 4 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure 02-S-01-4, Shift Relief and Turnover, Revision 42. Specifically, the licensee failed to ensure proper turnover of the status of temporary modifications installed in the plant was being conducted by operations staff during turnover. The inspectors determined that the operations staff was required by Attachment III of that procedure to review the TMs log prior to taking the shift. The inspectors interviewed the operations staff and asked if the TMs were reviewed prior to taking shift that day. The staff member stated he had not and when asked about Attachment III of the turnover procedure, he was not familiar with that attachment of the procedure. The inspectors interviewed additional operations staff members about the review of temporary modification status during turnover, and they also indicated they had not reviewed temporary modification during turnover. As a corrective action, the licensee added copies of Attachment III of the shift turnover procedure to the operations staff turnover book to ensure TMs were reviewed during shift turnover. The licensee entered this issue into the corrective action process under Condition Reports CR-GGN-2013-04481 and CR-GGN-2013-05955. The failure to review temporary modifications by operations personnel during turnover in accordance with station procedures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to more significant safety concerns. Specifically, operators not reviewing the status of TMs installed in the plant during turnover could result in a loss of configuration control of plant equipment that could result in an improper response by operators to plant events. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, Attachment 4, Table 3, the inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency affecting the design or qualification of a mitigating system, structure, or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allowed outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. The inspectors determined the apparent cause of this finding was that licensee personnel were not using Attachment III of the operations turnover procedure. Therefore, the finding has a cross-cutting aspect in human performance area associated with work practices in that the licensee management did not provide proper oversight to ensure a proper turnover was being conducted by operations personnel [H.4.(c)] (Section 1R18). Inspection Report# : 2013004 (pdf) Significance: Sep 30, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Maintain design Control of the Power Supplies for the Emergency Switchgear and Battery Room Fire Dampers Green. The inspectors reviewed a self-revealing Green non-cited violation of Facility Operating License Condition 2.C (41), Fire Protection Program, involving the failure to maintain design control of the power supplies for the emergency switchgear and battery room fire dampers. During a surveillance of the division 2 carbon dioxide Fire Damper Actuation System, ten division 1 switchgear and battery room cooler fire dampers were inadvertently closed. Electricians investigated and found that a common ground existed between the division 1 and 2 emergency switchgear and battery room damper control panels. The common ground was determined to originate from a factory installed ground strap connecting the negative terminal to the ground/neutral on the emergency switchgear and battery room damper control power supplies. The licensee reviewed plant drawings and determined that the ground strap on the power supplies should have been removed prior to installation due to this being designed as a non-grounded system. As an immediate corrective action, the licensee removed the factory installed ground straps and restored the system to operable status. The licensee entered this issue into the corrective action process under Condition Report CR-GGN-2013-03827. Page 5 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 The failure to verify a new power supply was a like-for-like replacement of the original power supply to ensure the replacement power supply did not alter the design of the damper control system was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, Attachment 4, Table 3, the inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The inspectors determined that the finding had an adverse effect on the fixed fire suppression systems. The inspectors assigned a low degradation rating due to the fact that the automatic fire suppression systems performance and reliability was minimally impacted by the inspection finding. Since the finding was assigned a low degradation rating, it screened as being of very low safety significance (Green). The apparent cause of this finding was the procurement engineering evaluation did not verify the replacement power supplies met the design requirements to be compatible with the unique design of the emergency switchgear and battery room damper control system. Therefore, the finding had a cross-cutting aspect in the area of human performance, work practices component because the licensee failed to properly perform a procurement evaluation in accordance with station procedures [H.4(b)] (Section 1R18). Inspection Report# : 2013004 (pdf) Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Implement a Compensatory Fire Watch per Station Fire Protection Procedures Green. The inspectors identified a non-cited violation of Facility Operating License Condition 2.C(41) for the failure to properly implement a compensatory fire watch per the station fire protection program. Following an inadvertent release of carbon dioxide from the Cardox automatic fire suppression system into a division 2 safety related switchgear room located in the auxiliary building, the operators isolated the auxiliary building from the Cardox system to prevent any future inadvertent releases. The inspectors accompanied the fire watch patrol, which was required due to the isolation of the Cardox system to the auxiliary building, and they noted that during the patrol, none of the 10 rooms requiring a fire watch were checked. The inspectors brought this concern to the shift manager who confirmed that each room was required to be checked per the established fire watch criteria and that the fire watch patrol misunderstood the requirement. The licensee took immediate corrective action to direct the fire watch to check all the rooms to restore compliance with the fire watch requirements. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-04058. The failure to perform a fire watch in accordance with the fire protection program is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it is associated with the protection against the external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to perform the fire watch correctly adversely impacted the plants capability to detect and suppress a fire in a timely manner. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, , Table 3, the inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The inspectors determined that the finding had an adverse affect on the fixed fire protection systems element of fire watches posted as a compensatory measure for outages or degradations. The inspectors assigned a high degradation rating due to the automatic fire suppression system being tagged out of service. Because the system was degraded without compensatory actions for less than three days, the Page 6 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 inspectors used a duration factor of 0.01. The inspectors used 2E-2 for a generic fire frequency area for a switchgear room. The resulting change in core damage frequency was 2E-4, which was greater than the high degradation Phase 1 Quantitative Screening Criteria of 1E-6. Therefore, a senior reactor analyst performed a detailed risk evaluation. The analyst performed a bounding analysis of the performance deficiency (See Table 1R05-1). For each of the 10 affected fire areas, the analyst determined the probability of a fire occurring by multiplying the fire ignition frequency from the licensees fire hazards analysis by the 9.2 hours that the performance deficiency impacted the plant. Because each fire area had a functional fire detection system throughout the exposure period, the analyst determined the non-detection probability by multiplying the fire probability by the generic failure probability for a detection system. The analyst made the bounding assumption that all fires postulated to initiate that were not detected would proceed to core damage. The sum of all the non-detection probabilities was 9.1 x 10-7 (See Table 1R05-1). Therefore, the bounding analysis indicates that this finding is of very low safety significance (Green). The inspectors determined the apparent cause of this finding was that the security officers performing the fire watch patrols did not understand the requirement to visually check the affected rooms. Therefore, the finding has a cross-cutting aspect in the human performance area associated with the work practices component because the licensee did not communicate human error prevention techniques such as pre-job briefings and proper documentation of activities commensurate with the risk of the assigned task [H.4(a)] (Section 1R05.1.b). Inspection Report# : 2013003 (pdf) Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Nonconforming Condition with the Standby Diesel Generator Inlet Plenum Turning Vanes Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. Specifically, from November 3, 1988, until March 6, 2010, actions to correct known design deficiencies on the left and right banks intercooler inlet plenums of both the division 1 and 2 standby diesel generators were not fully implemented. The design deficiency, identified by the vendor, could result in intercooler tube failure and jacket water leakage. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2013-02631. The failure to correct a nonconforming condition in the division 1 and 2 standby diesel generators inlet plenums is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it adversely affected the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage), and if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to implement corrective actions to resolve a known design deficiency of the intercooler inlet plenums could have resulted in either the division 1 or 2 standby diesel generator failing to perform its safety function. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green), because the finding was a design deficiency affecting a mitigating systems structure, system, or component that did not lose operability or functionality. The finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance (Section 1R17.1.b.2). Inspection Report# : 2013003 (pdf) Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Nonconforming Condition in the Train B Starting Circuit Page 7 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. Specifically, from November 20, 1998, until November 7, 2012, actions to correct a known nonconforming condition involving the low pressure interlock of the train B starting circuit on both the division 1 and 2 standby diesel generators had not been implemented. If the train A starting circuit were to fail and starting air pressure were to fall below 120 psig, the diesel generator would have all automatic shutdown permissives active, which are supposed to be bypassed during a loss-of-coolant-accident signal. This was considered a single point vulnerability for the train B starting circuit. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2013-02524. The failure to correct a nonconforming condition in the division 1 and 2 standby diesel generators train B starting circuits is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it affected the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage), and if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to implement corrective actions to resolve a known nonconforming condition of the low pressure interlock of the train B starting circuit could have resulted in either the division 1 or 2 standby diesel generator failing to perform its safety function. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green), because the finding was a design deficiency affecting a mitigating systems structure, system, or component that did not lose operability or functionality. The finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance (Section 1R17.1.b.3). Inspection Report# : 2013003 (pdf) Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Verify the Residual Heat Removal B System was Full of Water Within its Specified Frequency Green. The inspectors identified a non-cited violation of Technical Specification Surveillance Requirement SR 3.5.1.1 for the failure to verify the residual heat removal B system was full of water within its specified frequency. The inspectors reviewed a low pressure core injection subsystem B monthly functional test that was performed on April 10, 2013, per Procedure 06-OP-1E12-M-0002, LPCI/RHR Subsystem B Monthly Functional Test, Revision 113. The inspectors identified that the licensee failed to perform ultra sonic testing on the pipe prior to and after venting of the pipe directly upstream of the residual heat removal heat exchanger B outboard vent valve (1E12F074B). By not performing the ultra sonic testing, the licensee did not verify the residual heat removal system was full of water as required by Surveillance Requirement 3.5.1.1. Immediate corrective actions included performing the ultra sonic tests, which verified the system was full of water and satisfied the surveillance requirement. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2013-02847. The failure to verify the residual heat removal B system was full of water as required by Technical Specification Surveillance Requirement SR 3.5.1.1 is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstones objective of ensuring the availability, reliability and capability of systems that respond to prevent undesirable consequences. Specifically, the failure to perform the required ultra sonic testing resulted in Technical Specification Surveillance Requirement SR 3.5.1.1 not being met. Therefore, the licensee could not ensure the system would perform properly by injecting its full capacity into the reactor coolant system upon demand. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency Page 8 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 affecting the design or qualification of a mitigating system, structure, or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allowed outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. Through interviews with operations personnel, the inspectors determined the apparent cause of the finding was that management failed to ensure the ultra sonic test was performed. Therefore, the finding had a cross-cutting aspect in the human performance area associated with the work practices component because the licensee failed to ensure supervisory and management oversight of work activities [H.4(c)] (Section 1R22.b). Inspection Report# : 2013003 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Seal Safety-related Manholes Green. The inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, involving the failure to ensure that manhole MH01 was properly sealed to prevent entry of flammable liquid. Specifically, on February 20, 2013, four manhole covers had between one to three loose bolts and evidence of water seepage. These vaults contain safety related cables for standby service water trains A and B. Immediate corrective actions included cleaning and tapping the bolt holes to ensure proper thread engagement, adding work instructions to the preventative maintenance procedure to clean the manhole bolt holes, and verifying that the other manholes containing safety-related cables did not have similar issues with loose bolts on the manhole covers. The licensee entered this issue in their corrective action program as Condition Report CR GGN-2013-01348. This finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of protection against external factors and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and required the use of Inspection Manual Chapter 0609, , Appendix F, Fire Protection Significance Determination Process. However, an NRC senior reactor analyst determined that the unique nature of this performance deficiency did not lend itself to analysis by the methods provided in Appendix F. Therefore, a Phase 3 analysis was performed. Based on a bounding analysis, the analyst determined that the change in core damage frequency was approximately 1.5E-7/yr. The result was low because of the relatively short periods of time that fuel was actually being transferred, the low probability of transfer system failures, and the low likelihood that a loss of normal service water initiator would occur following a fire in the subject manholes. The finding has a cross-cutting aspect in the human performance area associated with the resources component because the licensee did not provide adequate work packages [H.2(c)] (Section 1R06). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Monitor for Ice on Standby Service Water Towers Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, for the licensees failure to monitor for ice accumulation on the standby service water cooling towers in accordance with station procedures. On January 17, 2013, the plant experienced a winter storm but operators did not implement Standby Service Water System Operating Instruction, 04-1-01-P41-1, Revision 137, Section 6.2, Cold Weather Operation, which directed the licensee to monitor the standby service water cooling tower for ice accumulation when weather conditions existed that could have resulted in icing of the cooling tower fill material and missile grating. The licensee Page 9 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 entered this issue into their corrective action program as Condition Report CR-GGNS-2013-00426. The failure to monitor for ice accumulation in accordance with station procedures is a performance deficiency. The finding is more than minor because if left uncorrected, it could lead to a more significant safety concern. Specifically, the occurrence of ice accumulation on the standby service water cooling towers, if unmonitored, could cause damage to the fill material and/or the tower missile gratings, which would render the standby service water system inoperable. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency affecting the design or qualification of a mitigating system, structure or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allow outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. The finding has a cross-cutting aspect in the human performance area associated with the work control component because the licensee failed to appropriately plan work activities based on environmental conditions that may impact plant structures, systems and components [H.3(a)] (Section 1R13). Inspection Report# : 2013002 (pdf) Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Design Control for Setpoint Calculations Green. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure of the licensee to maintain design control, incorporate, verify, and check new instrument drift values, and translate the design basis requirements for multiple allowable values and trip setpoints described in the technical specifications into setpoint calculations. During the review of condition reports associated with an operability review of the licensees transition from an 18- to 24-month operating cycle in August 2012, inspectors identified that the licensee failed to maintain design control of multiple setpoint calculations. In response to NRC inspector questioning, a licensee review of the calculations identified that three of the 14 calculations reviewed contained calculated allowable values that differed from the values contained in the Technical Specifications associated with Level 8 Narrow Range, Reactor Scram on High SDVP Water Level, and HPCS & RCIC Pump Suction Transfer on High Suppression Pool Level. An assessment of the calculations also determined that one other calculation contained an error that was introduced during the replacement of the high-pressure turbine rotor in a recent refueling outage, which would require a license amendment request. The licensee entered this condition in their corrective action program as CR-GGN-2013-00371. The failure to maintain design control, incorporate, verify, and check new instrument drift values, and translate the design basis requirements into multiple allowable values and trip setpoints described in the technical specifications into facility setpoint calculations is a performance deficiency. This finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of design control and affected the cornerstone objective of ensuring the capability of the safety-related system to respond to initiating events to prevent undesirable consequences. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the issue was determined to affect the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of the offsite power supply operability or functionality. This finding has a cross-cutting aspect in the area of human performance decision-making because the licensee did not use a systematic decision making process and did not obtain interdisciplinary input on a risk significant decision [H.1(a)] (Section 1R15). Inspection Report# : 2013002 (pdf) Page 10 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Scaffold Affecting Fire Brigade Access Green. The inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, for the failure to identify and correct a condition adverse to fire protection. Specifically, the licensee failed to ensure that fire brigade members had sufficient access through a scaffold built in the diesel generator building hallway into the division-1 diesel generator room. The immediate corrective actions included removing the scaffold in the diesel generator building hallway. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2013-01679. The failure to take prompt corrective action to ensure adequate access for fire brigade members through installed scaffolding in the diesel generator building hallway to the division-1 diesel generator room is a performance deficiency. The finding is more than because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the inability for fire brigade members to gain access to safety related equipment in timely manner could result in preventing prompt extinguishing of fires. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because the finding involved a risk-significant fire area that had an automatic fire suppression system. The inspectors determined the apparent cause of this finding was that the licensee did not implement the corrective action program with a low threshold for identifying scaffolding that could impede fire brigade member response during a fire. Therefore the finding had a cross-cutting aspect in the problem identification and resolution area associated with the corrective action program component because the licensee failed to identify conditions adverse to fire protection [P.1(a)] (Section 1R22). Inspection Report# : 2013002 (pdf) Significance: Feb 27, 2013 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Aligning Nitrogen Backup to Automatic Deressurization System

  • Green. The team identified a Green non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensees procedures for aligning portable nitrogen bottles to the Instrument Air system for backup operation of Automatic Depressurization System (ADS) valves do not include a step to direct the pressure regulator outlet isolation valves to be opened. If these valves are left closed, the nitrogen bottles will remain isolated from the Instrument Air system.

The failure to include a procedural step to open the nitrogen regulator outlet isolation valves when aligning nitrogen to the ADS valve instrument air lines is a performance deficiency. The performance deficiency is more than minor and is therefore a finding because it is associated with the procedure quality attribute of the mitigating systems cornerstone and affects the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This deficiency could have significantly affected the operators ability to perform the activity affecting quality, in this case, aligning nitrogen as a backup to ADS valve instrument air. Using Inspection Manual Chapter 0609, Attachment 4, Initial Screening and Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, a Phase 1 screening was performed and determined that the finding required a detailed risk evaluation because the finding would have resulted in a loss of system safety function had the procedure been called upon. Page 11 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 The senior reactor analyst performed a detailed risk evaluation using the Grand Gulf Standardized Plant Analysis Risk model Version 8.22, and the SPAR-H human reliability analysis method. This method resulted in an incremental conditional core damage probability of 7.0 x 10-6. However, the analyst noted that, given the specific performance deficiency, this method provided a bounding analysis. Therefore, the finding was assessed using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process using Qualitative Criteria. The analyst noted that licensee calculations and surveillance of the accumulators and associated check valves indicated that accumulator pressure would remain available for much longer than the 6 hours suggested in the model. Additionally, the failure of the 21 safety-relief valves under this condition would not occur simultaneously, but would be staggered as a result of the depressurization of individual accumulators. This would provide additional indication, cues, and time for operators to identify and correct the valve alignment error. Finally, the SPAR model does not consider the potential for recovery of the instrument air system. Based on this additional qualitative information, the analyst determined that the additional cues and time provided to the operators combined with the straight-forward diagnosis for this specific finding would reduce the overall risk of this performance deficiency by more than an order of magnitude. Therefore, using a bounding quantitative evaluation combined with qualitative factors, this finding was determined to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues by missing multiple opportunities to identify the procedural discrepancy when it was developing and validating the exam material for submission to the NRC [P.1(a)]. Inspection Report# : 2012301 (pdf) Barrier Integrity Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Comply with Technical Specification 3.4.11 The inspectors identified a non-cited violation of Technical Specification 3.4.11 for the failure to comply with the Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) during plant cold startups. Specifically, the PTLR had a lower limit of zero psig, and the licensee operated the reactor pressure vessel (RPV) below zero psig during the plant start-up that commenced on November 2, 2013. A review of plant data showed that the RPV pressure was maintained below zero psig for approximately 2 hours. The licensee performed an engineering evaluation and determined that the maximum compressive stress experienced by the RPV did not exceed the maximum yield strength of RPV. Immediate corrective action included revising Procedure 03-1-01-1, Cold Shutdown to Generator Carrying Minimum Load, to ensure the RPV is pressurized prior to opening the main steam isolation valves (MSIVs) and providing training on the procedural changes to all the operating crews. The licensee entered this issue into the corrective action process under Condition Report CR-GGN-2013-07021. The failure to comply with the RCS Pressure and Temperature Limits Report specified in Technical Specification 3.4.11 was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and had the potential to adversely affect the associated cornerstone objective of providing reasonable assurance that a physical design barrier (reactor coolant system) protects the public from radionuclide release caused by accidents or events. Specifically, without NRC review and approval of revised pressure and temperature limits that include operating the RPV below zero psig, the inspectors did not have reasonable assurance the RPV would not be Page 12 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 adversely affected. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," June 19, 2012, the inspectors determined that the issue affected the Barrier Integrity Cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, June 19, 2012, Exhibit 3, the inspectors determined that since this finding involved the reactor coolant system boundary, a detailed risk evaluation was required. The Senior Reactor Analyst reviewed the finding and determined that a detailed risk evaluation was not required. The licensee performed an engineering evaluation and concluded that there was no impact to the reactor vessel. As a result, the Senior Reactor Analyst concluded that there was no change in risk due to the performance deficiency. The inspectors determined that since the procedural steps to perform Attachments VIII and X concurrently had been in place since 1994, this was a latent issue; therefore no cross-cutting aspect was assigned. Inspection Report# : 2013005 (pdf) Significance: Dec 31, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Provide Adequate Procedures Results in Loss of Safety Function The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to provide an adequate procedure for a safety related activity. On December 17, 2013, while performing Surveillance Procedure 06-IC-1E31-Q-1016-02, RCIC Steam Supply Pressure Low Functional Test, Revision 111, the reactor core isolation cooling (RCIC) system became inoperable due to the procedure being incorrectly revised. Furthermore, the procedure error resulted in the containment isolation capability for RCIC being lost for approximately 1 hour. As an immediate corrective action, the licensee restored the breakers regaining isolation capability, and reopened the RCIC inboard isolation valve, thus restoring RCIC to operable status. The licensee entered this issue into the corrective action process under Condition Reports CR-GGN-2013-07720, CR-GGN-2013-07733, and CR-GGN-2013-07374. The failure to have an adequate procedure for the reactor core isolation cooling steam supply pressure low functional test is a performance deficiency. The performance deficiency was more than minor and therefore a finding because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This performance deficiency was also associated with the procedural quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, the inspectors determined the issue affected the Barrier Integrity Cornerstone. The inspectors used Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, May 6, 2004, and determined the finding was a type B finding at full power. Using Table 6.1, Phase 1 Screening-Type B Findings at Power, the inspectors concluded that since this issue involved containment isolation valves in a BWR Mark III containment, a Phase 2 analysis was necessary. Using Table 6.2, Phase 2 Risk Significance - Type B Findings at Full Power, the inspectors concluded that the risk significance was very low (Green) because the exposure time was less than 3 days. Furthermore, the inspectors determined that this issue affected the Mitigating System Cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, June 19, 2012, Exhibit 2, the inspectors determined that since the finding represented a loss of system and/or function, a detailed risk evaluation was required. The inspectors utilized the Grand Gulf Standardized Plant Analysis Risk model to determine the change in core damage frequency (CDF) due to the loss of safety function. The inspectors assigned the RCIC system a failure probability of 1.00 for a conservative duration of 1 hour. The resulting change in CDF was 1.9E-9/year, thus the finding was of very low safety significance (Green). The Senior Risk Analyst reviewed the inspectors evaluation and verified the conclusions to be correct. The apparent cause of this finding was that the licensee failed to effectively utilize human error prevention techniques. Therefore, the finding had a cross-cutting aspect in the area of human performance, work practices because the licensee did not perform adequate Page 13 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 self and peer checking while performing an activity affecting quality [H.4(a)] Inspection Report# : 2013005 (pdf) Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Obtain NRC Approval for a Change in Method of Evaluation for Determining Reactor Vessel Fluence SL-IV. The team identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, involving the licensees failure to obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a new method of evaluation for determining reactor vessel neutron fluence. On November 4, 2003, the NRC issued Amendment Number 160 to the Facility Operating License of the Grand Gulf Nuclear Station. The amendment revised the Updated Final Safety Analysis Report (UFSAR) to change the Reactor Vessel Material Surveillance Program to reflect participation in the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). Additionally, the amendment revised the UFSAR to state that neutron fluence calculations performed after 2002 will be in accordance a methodology that has been approved by the NRC staff and is consistent with the attributes identified in NRC Regulatory Guide 1.190, Calculation and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. The licensee developed a new neutron fluence calculation method which was based on a neutron fluence calculation method that had been previously approved by the NRC for another facility, which was documented in Nine Mile Point Nuclear Station, Unit No. 1 - Issuance of Amendment RE: Pressure-Temperature Limit Curves and Tables, dated October 27, 2003. The NRC identified that the calculation, which was developed for GGNS, used the CASMO-4/SIMULATE code package to calculate the neutron source, whereas the prior calculation performed for Nine Mile Point Nuclear Station (NMP) used the ORIGEN code to calculate the neutron source. The inspectors determined that, although these codes are intended for the same purpose, they are distinct codes and the NRC approved only the use of one neutron source code (i.e., ORIGEN) in the neutron fluence calculation method of evaluation at Nine Mile Point. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2013-04743. The licensees failure to determine that a change to their method of evaluation for calculating reactor vessel neutron fluence was a departure from a method of evaluation approved by the NRC and required NRC review and approval prior to implementation was a performance deficiency. The performance deficiency was evaluated using traditional enforcement because the finding had the ability to impact the regulatory process. The performance deficiency was more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation. In accordance with the NRC Enforcement Manual, risk insights from Inspection Manual Chapter 0609, Significance Determination Process, are used in determining the significance of 10 CFR 50.59 violations. Using the Inspection Manual Chapter 0612, Appendix B, Issue Screening, the team determined the finding adversely affected the Barrier Integrity Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the team determined the finding required a detailed risk evaluation because the finding involved the reactor coolant system boundary. A Senior Reactor Analyst performed the evaluation and determined the finding had very low safety significance (i.e., Green) because the NRC performed calculations and did not determine that the licensees Pressure-Temperature limits had or would have expired or been invalid; therefore, the change in risk was negligible. Since the finding had very low safety significance, the finding was determined to be Severity Level IV, in accordance with the NRC Enforcement Policy. The finding does not have a cross-cutting aspect because cross-cutting aspects are not assigned to traditional enforcement violations (Section 1R17). Inspection Report# : 2013004 (pdf) Significance: N/A Jun 28, 2013 Page 14 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Identified By: NRC Item Type: NCV NonCited Violation Failure to Revise Figures and Tables in the Updated Final Safety Analysis Report SLIV. The inspectors identified a non-cited violation of 10 CFR 50.71(e)(4), which requires the Final Safety Analysis Report be updated, at intervals not exceeding 24 months, and states in part, the revisions must reflect all changes made in the facility or procedures described in the FSAR. Specifically, the inspectors identified three examples of changes to figures or tables that had not been included in the licensees Updated Final Safety Analysis Report submittal in November, 2012: (1) Figure 9.2-027, Sheet 2, Revision 16 Plant Service Water Radial Well System Unit 1 (2) Figure 10.4-011, Condensate System (Drawing M-1053B, Revision 28), and (3) Table 9.1-12, Maximum Fuel Pool Heat Load did not include values associated with the extended power uprate. This finding has been entered into the licensees corrective action program as Condition Reports CR-GGN-2013-00426, CR-GGN-2013-02661, and CR-GGN-2013-02471. The failure of the licensee to include all changes made to the facility or procedures in their November 2012 update to the original revision of the Final Safety Analysis Report is a performance deficiency. The issue is a performance deficiency because it was a failure to meet a requirement, 10 CFR 50.71(e)(4), and it was within the licensees ability to correct this problem. Using Inspection Manual Chapter 0612, Appendix B, the performance deficiency was assessed through both the Reactor Oversight Process and traditional enforcement because the finding had the potential for impacting the NRCs ability to perform its regulatory function. By screening through the Reactor Oversight Process, the finding resulted in a minor performance deficiency. Following the traditional enforcement path, the inspectors used the NRC Enforcement Policy, dated January 28, 2013, to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy and in accordance with Section 6.1.d.3, this finding was determined to be a Severity Level IV non-cited violation because the licensee failed to update the Final Safety Analysis Report as required by 10 CFR 50.71(e)(4). However, the lack of up-to-date information had not resulted in any unacceptable change to the facility or procedures. This finding had no cross-cutting aspect (Section 1R17.1.b.1). Inspection Report# : 2013003 (pdf) Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Entry Into A High Radiation Area Without A Required Radiation Monitoring Device The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.7.1, resulting from an individual entering a high radiation area without the required radiation monitoring device. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2012-04112. As a corrective action, the radiation protection manager coached the individual on the need for proper dosimetry devices in high radiation areas. The entry into a high radiation area without all required radiation monitoring devices was a performance deficiency and was a violation of Technical Specification 5.7.1. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because it removed a barrier intended to prevent the worker from receiving unexpected dose. Page 15 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation had a cross-cutting aspect in the human performance area, associated with the work practices component, because the worker and crew members did not use human error prevention techniques, such as self and peer checking [H.4(a)]. Inspection Report# : 2013005 (pdf) Significance: Dec 31, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Survey Resulting in Personnel Entry To A High Radiation Area The inspectors reviewed a self-revealing, non-cited violation of 10 CFR 20.1501(a) for failure to survey, which resulted in a worker entering an unposted high radiation area. This issue was entered into the licensees corrective action program as Condition Reports CR-GGN-2012-08436 and CR-GGN-2012-09225. As corrective actions, the licensee coached radiation protection personnel on exhibiting a questioning attitude, walked down all affected areas; verified correct postings were used, and surveyed for any other unanticipated dose rate alarms. The failure to survey and determine radiation levels was a performance deficiency. The significance of the performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because the failure exposed a pipefitter to higher than anticipated radiation dose rates. The inspectors used Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, to determine the significance of the violation. The violation had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation had a cross-cutting aspect in the human performance area, associated with the work control component, because licensee personnel failed to appropriately plan a work activity by not incorporating risk insights, job site conditions, including environmental conditions, which may impact human system interface and radiological safety, and the need for planned contingencies or compensatory actions, such as surveying and up-posting affected areas after a power ascension [H.3(a)]. Inspection Report# : 2013005 (pdf) Public Radiation Safety Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement the Offsite Dose Calculation Manual Green. Inspectors identified three examples of a non-cited violation of Technical Specification 5.5, Programs and Manuals, for failure to maintain and implement requirements of the offsite dose calculation manual (ODCM). Specifically, the licensee failed to: (1) adequately document and justify ODCM changes, (2) approve licensee initiated changes to the ODCM, and (3) implement the radiological effluent controls for liquid releases. The violation was entered into the licensees corrective action program as Condition Report CR-GGN-2013-05039, and the licensee is evaluating the issue to determine the proper corrective action. Failure to implement the requirements of the offsite dose calculation manual is a performance deficiency. This Page 16 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 performance deficiency is more than minor because it affected the Public Radiation Safety Cornerstone attribute of program and process because the failure to adequately justify and approve offsite dose calculation manual changes resulted in 49 liquid effluent releases, contrary to the licensees Offsite Dose Calculation Manual, Revision 37, requirements. Using Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, the inspectors determined this to be a violation of very low safety significance (Green). The violation was in the effluent release program but was not a substantial failure to implement the effluent program, and the dose to the public did not exceed the 10 CFR Part 50 Appendix I criterion or 10 CFR 20.1301(e) limits. The violation had a cross-cutting aspect in the human performance area associated with the resources component because the licensee failed to ensure the individuals preparing and reviewing offsite dose calculation manual changes had sufficient knowledge of the effluent release control system, its components, and its function to adequately evaluate the impact of the change [H.2(b)] (Section 2RS6). Inspection Report# : 2013004 (pdf) Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Include Some Solid Radwaste Released in the 2012 Regulatory Guide 1.21 Annual Effluent Report Green. Inspectors identified a non-cited violation of Technical Specification 5.6.3 because the licensee failed to include in the 2012 Annual Radiological Effluent Release Report some solid radioactive waste released to an offsite waste processor. The failure to include in the 2012 Annual Radiological Effluent Release Report all solid radioactive waste released to an offsite waste processor was a performance deficiency, contrary to Technical Specification 5.6.3. The violation was determined to be more than minor because it was associated with the Public Radiation Safety Cornerstone attribute of program and process and adversely 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, in that some licensed radioactive material, which left the Grand Gulf Nuclear Station, was unaccounted for. Using Inspection Manual Chapter 0609, Appendix D, "Public Radiation Safety Significance Determination Process," dated February 12, 2008, the inspectors determined the violation to be of very low safety significance because, although it was a radioactive material control issue, it was not a transportation issue, and it did not result in public dose greater than 0.005 rem. The violation had a cross-cutting aspect in the human performance area, work control component because the licensee did not appropriately coordinate work activities by incorporating actions to address the need for work groups to communicate and coordinate with each other during activities in which interdepartmental coordination was necessary to assure human performance [H.3(b)] (Section 2RS8). Inspection Report# : 2013004 (pdf) Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary. Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed. Page 17 of 18

4Q/2013 Inspection Findings - Grand Gulf 1 Miscellaneous Significance: N/A Dec 05, 2013 Identified By: NRC Item Type: FIN Finding Grand Gulf 2013 Biennial Problem Identification and Resolution Inspection Summary The team reviewed a sample of system health reports, self assessments, trending reports and metrics, and various other documents related to the corrective action program. Licensee identified problems were entered into the corrective action program at a low threshold. Problems were generally prioritized and evaluated commensurate with the safety significance of the problems and corrective actions were generally implemented in a timely manner. Corrective actions were generally implemented in a timely manner commensurate with their importance to safety and addressed the identified causes of problems. The licensee appropriately evaluated industry operating experience for relevance to the facility and had entered applicable items in the corrective action program. The licensee used industry operating experience when performing root cause and apparent cause evaluations. The licensee performed effective quality assurance audits and self assessments, as demonstrated by self identification of poor corrective action program performance and identification of ineffective corrective actions. Inspection Report# : 2013007 (pdf) Significance: N/A May 25, 2012 Identified By: NRC Item Type: VIO Violation Violation for Grand Gulf (2012 Findings) Regulatory requirement: 10 CFR 54.13(a) states, in part, that information provided to the Commission by an applicant for a renewed lic3ense must be complete and accurate in all material respects. Apparent violation: Contrary to the above, Entergy Operations, Inc (EOI) provided information to the NRC, for a renewed license at the Grand Gulf Nuclear Station (GGNS), in responses to several requests for additional information (RAIs) that was not complete and accurate in all material respects. The inaccurate information in the RAI responses was material to the NRC because the NRC relies on the information in RAI responses to determine whether the licensee has demonstrated that aging effects will be adequately managed as required by 10 CFR 54.21(a)(3). Inspection Report# : 2013201 (pdf) Last modified : February 24, 2014 Page 18 of 18

1Q/2014 Inspection Findings - Grand Gulf 1 Grand Gulf 1 1Q/2014 Plant Inspection Findings Initiating Events Significance: Sep 30, 2013 Identified By: NRC Item Type: FIN Finding Failure to Follow Alarm Response Steps to Restore the TSE Following Maintenance The inspectors reviewed a Green self-revealing finding for the failure to follow Procedure 04-1-02-1H13-P680-9A, TSE INFL OFF, Revision 36; in that operations personnel did not verify steps were followed per this alarm response procedure prior to returning the turbine thermal stress evaluator (TSE) to service following maintenance activities. The failure to follow alarm response procedure then resulted in an automatic reactor scram on July 30, 2013. Site personnel determined that the scram was caused by high reactor pressure resulting from the turbine unloading beyond the capability of the bypass valves after restoring the TSE to service following maintenance. On July 26, 2013, the control room received an alarm "TSE-STU CAB FAIL." The licensee failed to determine the correct cause of the alarm due to inadequate troubleshooting. Therefore, when the maintenance was completed and the TSE was returned to service, the turbine started to unload resulting in a reactor scram due to reactor vessel high pressure. The immediate corrective actions included determining the cause of the scram and taking actions to restore equipment prior to plant startup. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2013-04943. The failure to follow alarm response steps to restore the TSE following maintenance is a performance deficiency. Specifically, Procedure 04-1-02-1H13-P680-9A, TSE INFL OFF, Revision 36, step 4.1 requires operational personnel to ensure that the TSE is functioning correctly following maintenance prior to restoring to service. The performance deficiency is more than minor, and therefore a finding, because it is associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Initiating Events Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that the issue has a very low safety significance (Green) because it only caused a reactor trip and did not cause a loss of mitigating equipment relied on to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined that the apparent cause of the finding was that the licensee did not troubleshoot to validate the cause for alarm TSE STU Cab Failure in accordance with station troubleshooting procedures. Therefore, the finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee did not use the troubleshooting process effectively [H.4(b)]. Inspection Report# : 2013004 (pdf) Mitigating Systems Significance: Mar 31, 2014 Page 1 of 14

1Q/2014 Inspection Findings - Grand Gulf 1 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Scaffold Activity Would not Interfere with Fire Brigade Response The inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, for the failure to adhere to procedural requirements to ensure that scaffold installed in the plant would not prevent or restrict the fire brigade from accessing a certain route used for response to a fire in the area. On February 4, 2014, the licensee installed a scaffold in the containment building for an inspection. The licensees procedure required a walkdown of proposed scaffold to determine if the scaffold would prevent or restrict fire brigade access. The initial reviewer identified that the ladder to access the scaffold would restrict fire brigade access, thus the ladder was not installed until it was required. On March 1, 2014, the ladder was installed for the four hour inspection. Once completed, the licensee failed to remove the scaffold ladder to restore normal access to the area. On March 4, 2014, the inspectors identified that the scaffold ladder was still installed. The inspectors brought their concern to the licensee, who determined that the scaffold would adversely affect the response of fire brigade members to that area of containment. As an immediate corrective action, the licensee removed the scaffold ladder to allow adequate access for the fire brigade members. The licensee documented this issue in Condition Report CR-GGN-2014-02363. The failure to ensure fire brigade members had adequate access passed a scaffold installed in the containment building was a performance deficiency. The performance deficiency was more than minor and therefore a finding because it adversely impacted the protection against external factors attribute of the Mitigating System Cornerstone in that the fire brigades inability to gain access to certain areas in containment could result in preventing prompt extinguishing of fires. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and that the finding pertained to a degraded condition while the plant was shutdown for refueling outage RF19. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, dated February 28, 2005. The inspectors determined that Appendix G did not address fire brigade issues and solicited input from the senior reactor analyst. The senior reactor analyst performed a detailed risk evaluation and determined that Inspection Manual 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, June 19, 2012, Exhibit 2, Mitigating System Screening Questions, adequately bounded the performance deficiency. The inspectors determined that the finding involved the response time of the fire brigade to a fire, and the finding was of very low safety consequence (Green) because the fire brigades response time was mitigated by other defense-in-depth elements such as area combustible limits were not exceeded, installed fire detection systems were functional, and alternate means of safe shutdown were not impacted. Specifically, there were no combustibles in the area beyond limits, all fire detectors for the area were functional, and the plant was in a shutdown condition with the cavity flooded at the time. The apparent cause of this finding was the work groups involved did not communicate the significance of the impact the scaffold ladder had on fire brigade access to the area and the importance of having the ladder removed upon completion of the work. Therefore, the finding has a cross-cutting aspect in the human performance area associated with team work, in that the individuals and workgroups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Inspection Report# : 2014002 (pdf) Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow Procedure Results in Inadequate Operability Determination The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure EN-OP-104, Operability Determinations. Specifically, the inspectors identified that the licensee failed to establish an adequate basis for operability when a degraded or nonconforming condition had been identified. On August 30, 2013, Condition Report CR-GGN-2013-05604 was initiated to document a step change in the standby service water (SSW) siphon line K factor, which is a measure of flow through the siphon line. The K factor could have increased due to air Page 2 of 14

1Q/2014 Inspection Findings - Grand Gulf 1 entrapment in the siphon line that resulted from using air to mix the basin water following chemical treatments. The inspectors challenged the validity of the evaluation because the second step change in K factor, from 48 to 64, represented new information that had not been evaluated in the previous condition report. As an immediate corrective action, the licensee re-performed the operability determination and provided an adequate basis of operability by evaluating the system with the additional K factor data. Furthermore, the licensee verified the siphon line did not have any obstructions by observing the SSW basin levels equalize as water flowed through the siphon line. The licensee entered this issue into the corrective action process under Condition Report CR-GGN-2013-05687. The failure to perform an operability determination in accordance with procedure was a performance deficiency. The performance deficiency was more than minor, and is therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the reliability, availability and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the issue has very low safety significance (Green) because all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. The inspectors determined that the apparent cause of this finding was that the licensee had identified and used previously completed operability evaluations without verifying that the previously completed evaluations were fully applicable to the identified conditions. Therefore, the finding had a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly evaluate for operability conditions adverse to quality [P.1(c)]. Inspection Report# : 2013004 (pdf) Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Review Temporary Modifications by Operations Personnel During Turnover The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, regarding the licensees failure to follow the requirements of Procedure 02-S-01-4, Shift Relief and Turnover, Revision 42. Specifically, the licensee failed to ensure proper turnover of the status of temporary modifications installed in the plant was being conducted by operations staff during turnover. The inspectors determined that the operations staff was required by Attachment III of that procedure to review the TMs log prior to taking the shift. The inspectors interviewed the operations staff and asked if the TMs were reviewed prior to taking shift that day. The staff member stated he had not and when asked about Attachment III of the turnover procedure, he was not familiar with that attachment of the procedure. The inspectors interviewed additional operations staff members about the review of temporary modification status during turnover, and they also indicated they had not reviewed temporary modification during turnover. As a corrective action, the licensee added copies of Attachment III of the shift turnover procedure to the operations staff turnover book to ensure TMs were reviewed during shift turnover. The licensee entered this issue into the corrective action process under Condition Reports CR-GGN-2013-04481 and CR-GGN-2013-05955. The failure to review temporary modifications by operations personnel during turnover in accordance with station procedures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to more significant safety concerns. Specifically, operators not reviewing the status of TMs installed in the plant during turnover could result in a loss of configuration control of plant equipment that could result in an improper response by operators to plant events. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, Attachment 4, Table 3, the Page 3 of 14

1Q/2014 Inspection Findings - Grand Gulf 1 inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that the issue had a very low safety significance (Green) because it was not a deficiency affecting the design or qualification of a mitigating system, structure, or component, does not represent a loss of system or function, does not represent a loss of function for greater than its technical specification allowed outage time, and does not represent a loss of function as defined by the licensees Maintenance Rule program for greater than 24 hours. The inspectors determined the apparent cause of this finding was that licensee personnel were not using Attachment III of the operations turnover procedure. Therefore, the finding has a cross-cutting aspect in human performance area associated with work practices in that the licensee management did not provide proper oversight to ensure a proper turnover was being conducted by operations personnel [H.4.(c)]. Inspection Report# : 2013004 (pdf) Significance: Sep 30, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure to Maintain design Control of the Power Supplies for the Emergency Switchgear and Battery Room Fire Dampers The inspectors reviewed a self-revealing Green non-cited violation of Facility Operating License Condition 2.C (41), Fire Protection Program, involving the failure to maintain design control of the power supplies for the emergency switchgear and battery room fire dampers. During a surveillance of the division 2 carbon dioxide Fire Damper Actuation System, ten division 1 switchgear and battery room cooler fire dampers were inadvertently closed. Electricians investigated and found that a common ground existed between the division 1 and 2 emergency switchgear and battery room damper control panels. The common ground was determined to originate from a factory installed ground strap connecting the negative terminal to the ground/neutral on the emergency switchgear and battery room damper control power supplies. The licensee reviewed plant drawings and determined that the ground strap on the power supplies should have been removed prior to installation due to this being designed as a non-grounded system. As an immediate corrective action, the licensee removed the factory installed ground straps and restored the system to operable status. The licensee entered this issue into the corrective action process under Condition Report CR-GGN-2013-03827. The failure to verify a new power supply was a like-for-like replacement of the original power supply to ensure the replacement power supply did not alter the design of the damper control system was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, Attachment 4, Table 3, the inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The inspectors determined that the finding had an adverse effect on the fixed fire suppression systems. The inspectors assigned a low degradation rating due to the fact that the automatic fire suppression systems performance and reliability was minimally impacted by the inspection finding. Since the finding was assigned a low degradation rating, it screened as being of very low safety significance (Green). The apparent cause of this finding was the procurement engineering evaluation did not verify the replacement power supplies met the design requirements to be compatible with the unique design of the emergency switchgear and battery room damper control system. Therefore, the finding had a cross-cutting aspect in the area of human performance, work practices component because the licensee failed to properly perform a procurement evaluation in accordance with station procedures [H.4(b)]. Inspection Report# : 2013004 (pdf) Page 4 of 14

1Q/2014 Inspection Findings - Grand Gulf 1 Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Implement a Compensatory Fire Watch per Station Fire Protection Procedures The inspectors identified a non-cited violation of Facility Operating License Condition 2.C(41) for the failure to properly implement a compensatory fire watch per the station fire protection program. Following an inadvertent release of carbon dioxide from the Cardox automatic fire suppression system into a division 2 safety related switchgear room located in the auxiliary building, the operators isolated the auxiliary building from the Cardox system to prevent any future inadvertent releases. The inspectors accompanied the fire watch patrol, which was required due to the isolation of the Cardox system to the auxiliary building, and they noted that during the patrol, none of the 10 rooms requiring a fire watch were checked. The inspectors brought this concern to the shift manager who confirmed that each room was required to be checked per the established fire watch criteria and that the fire watch patrol misunderstood the requirement. The licensee took immediate corrective action to direct the fire watch to check all the rooms to restore compliance with the fire watch requirements. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2013-04058. The failure to perform a fire watch in accordance with the fire protection program is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it is associated with the protection against the external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to perform the fire watch correctly adversely impacted the plants capability to detect and suppress a fire in a timely manner. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," the inspectors determined that the issue affected the Mitigating Systems Cornerstone. Using NRC Inspection Manual Chapter 0609, , Table 3, the inspectors were directed to NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. The inspectors determined that the finding had an adverse affect on the fixed fire protection systems element of fire watches posted as a compensatory measure for outages or degradations. The inspectors assigned a high degradation rating due to the automatic fire suppression system being tagged out of service. Because the system was degraded without compensatory actions for less than three days, the inspectors used a duration factor of 0.01. The inspectors used 2E-2 for a generic fire frequency area for a switchgear room. The resulting change in core damage frequency was 2E-4, which was greater than the high degradation Phase 1 Quantitative Screening Criteria of 1E-6. Therefore, a senior reactor analyst performed a detailed risk evaluation. The analyst performed a bounding analysis of the performance deficiency (See Table 1R05-1). For each of the 10 affected fire areas, the analyst determined the probability of a fire occurring by multiplying the fire ignition frequency from the licensees fire hazards analysis by the 9.2 hours that the performance deficiency impacted the plant. Because each fire area had a functional fire detection system throughout the exposure period, the analyst determined the non-detection probability by multiplying the fire probability by the generic failure probability for a detection system. The analyst made the bounding assumption that all fires postulated to initiate that were not detected would proceed to core damage. The sum of all the non-detection probabilities was 9.1 x 10-7 (See Table 1R05-1). Therefore, the bounding analysis indicates that this finding is of very low safety significance (Green). The inspectors determined the apparent cause of this finding was that the security officers performing the fire watch patrols did not understand the requirement to visually check the affected rooms. Therefore, the finding has a cross-cutting aspect in the human performance area associated with the work practices component because the licensee did not communicate human error prevention techniques such as pre-job briefings and proper documentation of activities commensurate with the risk of the assigned task [H.4(a)]. Inspection Report# : 2013003 (pdf) Significance: Jun 28, 2013 Identified By: NRC Page 5 of 14

1Q/2014 Inspection Findings - Grand Gulf 1 Item Type: NCV NonCited Violation Failure to Correct a Nonconforming Condition with the Standby Diesel Generator Inlet Plenum Turning Vanes The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. Specifically, from November 3, 1988, until March 6, 2010, actions to correct known design deficiencies on the left and right banks intercooler inlet plenums of both the division 1 and 2 standby diesel generators were not fully implemented. The design deficiency, identified by the vendor, could result in intercooler tube failure and jacket water leakage. The finding was entered into the licensees corrective action program as Condition Report CR-GGN-2013-02631. The failure to correct a nonconforming condition in the division 1 and 2 standby diesel generators inlet plenums is a performance deficiency. The performance deficiency is more than minor and therefore a finding because it adversely affected the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage), and if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to implement corrective actions to resolve a known design deficiency of the intercooler inlet plenums could have resulted in either the division 1 or 2 standby diesel generator failing to perform its safety function. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green), because the finding was a design deficiency affecting a mitigating systems structure, system, or component that did not lose operability or functionality. The finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance. Inspection Report# : 2013003 (pdf) Significance: Jun 28, 2013 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct a Nonconforming Condition in the Train B Starting Circuit The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. Specifically, from November 20, 1998, until November 7, 2012, actions to correct a known nonconforming condition involving the low pressure interlock of the train B starting circuit on both the division 1 and 2 standby diesel generators had not been implemented. If the train A starting circuit were to fail and starting air pressure were to fall below 120 psig, the diesel generator would have all automatic shutdown permissives active, which are supposed to be bypassed during a loss-of-coolant-accident signal. This was considered a]]

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