IR 05000416/2011004

IR 05000416-11-004; 06/28 - 09/27/2011; Grand Gulf Nuclear Station, Integrated Resident Report and Regional Report; Operability Evaluations and Identification and Resolution of Problems

ML113140185
Person / Time
Site:	Grand Gulf
Issue date:	11/09/2011
From:	Vincent Gaddy NRC/RGN-IV/DRP/RPB-C
To:	Mike Perito Entergy Operations
References
IR-11-004
Download: ML113140185 (57)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGI ON I V 612 EAST LAMAR BLVD, SUITE 400 ARLINGTON, TEXAS 76011-4125 November 9, 2011 Mike Perito Vice President Operations Entergy Operations, Inc.

Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 Subject: GRAND GULF - NRC INTEGRATED INSPECTION REPORT NUMBER 05000416/2011004

Dear Mr. Perito:

On September 27, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Grand Gulf Nuclear Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on October 11, 2011, with you and other members of your staff.

The inspections examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, the NRC has identified three issues that were evaluated under the risk significance determination process as having very low safety significance (Green). The NRC has determined that violations are associated with these issues.

Additionally, one licensee-identified violation, which was determined to be of very low safety significance, is listed in this report. However, because of the very low safety significance and because they were entered into your corrective action program, the NRC is treating these findings as a noncited violations, consistent with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the significance of the noncited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E.

Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the facility. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this

Entergy Operations, Inc. -2-inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector at the facility.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response, if you choose to provide one, will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy or proprietary information so that it can be made available to the Public without redaction.

Sincerely, RC Hagar for VGaddy Vincent Gaddy, Chief Project Branch C Division of Reactor Projects Docket: 50-416 License: NPF-29 Enclosure:

NRC Inspection Report 05000416/2011004 w/Attachment: Supplemental Information Distribution via Listserv for GGNS

Entergy Operations, Inc. -3-Electronic distribution by RIV:

Regional Administrator (Elmo.Collins@nrc.gov)

Deputy Regional Administrator (Art.Howell@nrc.gov)

DRP Director (Kriss.Kennedy@nrc.gov)

DRP Deputy Director (Troy.Pruett@nrc.gov)

DRS Director (Anton.Vegel@nrc.gov)

DRS Deputy Director (Tom.Blount@nrc.gov)

Senior Resident Inspector (Rich.Smith@nrc.gov)

Resident Inspector (Blake.Rice@nrc.gov)

Branch Chief, DRP/C (Vincent.Gaddy@nrc.gov)

Senior Project Engineer, DRP/C (Bob.Hagar@nrc.gov)

Project Engineer, DRP/C (Rayomand.Kumana@nrc.gov)

Project Engineer, DRP/C (Jonathan.Braisted@nrc.gov)

GG Administrative Assistant (Alley.Farrell@nrc.gov)

Public Affairs Officer (Victor.Dricks@nrc.gov)

Public Affairs Officer (Lara.Uselding@nrc.gov)

Project Manager (Alan.Wang@nrc.gov)

Acting Branch Chief, DRS/TSB (Dale.Powers@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov)

Regional Counsel (Karla.Fuller@nrc.gov)

Congressional Affairs Officer (Jenny.Weil@nrc.gov)

RIV/ETA: OEDO (Mark.Franke@nrc.gov)

DRS/TSB STA (Dale.Powers@nrc.gov)

OEMail Resource ROPreports File located: R:\_REACTORS\_GG\2011\GG 2011004 RP-RLS.docx SUNSI Rev Compl. Yes No ADAMS Yes No Reviewer Initials VGG Publicly Avail Yes No Sensitive Yes No Sens. Type Initials VGG SRI:DRP/PBC RI:DRP/PBC SPE:DRP/PBC C:DRS/EB1 C:DRS/EB2 RLSmith BBRice BHagar TRFarnholtz NFOKeefe E-RLS FOR RCH E-RLS FOR /RA/ RLM FOR VGG /RA/

RCH 11/9/11 11/9/11 11/8/11 10/26/11 10/26/11 C:DRS/OB AC:TSS C:DRS/PSB1 C:DRS/PSB2 C:DRP/C MHaire DPowers MHay GEWerner VGaddy

/RA/ /RA/ /RA/ /RA/ RCH FOR VGG 10/21/11 11/9/11 11/8/11 11/8/11 11/9/11 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000416 License: NPF-29 Report: 05000416/2011004 Licensee: Entergy Operations, Inc.

Facility: Grand Gulf Nuclear Station Location: 7003 Baldhill Road Port Gibson, MS 39150 Dates: June 28, 2011, through September 27, 2011 Inspectors: R. Smith, Senior Resident Inspector B. Rice, Resident Inspector B. Baca, Health Physicist L. Carson II, Senior Health Physicist N. Greene, Ph.D., Health Physicist B. Larson, Senior Operations Engineer C. Steely, Operations Engineer Approved By: Vincent Gaddy, Chief Reactor Project Branch C Division of Reactor Projects-1- Enclosure

SUMMARY OF FINDINGS

IR 05000416/2011004; 06/28 - 09/27/2011; Grand Gulf Nuclear Station, Integrated Resident

Report and Regional Report; Operability Evaluations and Identification and Resolution of Problems.

The report covered a 3-month period of inspection by resident inspectors and two announced baseline inspections by regional inspectors. Three Green, noncited violations of significance were identified. The significance of most findings is indicated by their color (Green, White,

Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination Process.

The cross-cutting aspect is determined using Inspection Manual Chapter 0310, Components Within the Cross Cutting Areas. Findings for which the significance determination process does not apply may be Green or may be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

Cornerstone: Barrier Integrity

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).

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).

Cornerstone: Mitigating Systems

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).

Licensee-Identified Violations

A violation of very low safety significance, which was identified by the licensee, has been reviewed by the inspectors. Corrective actions, taken or planned by the licensee, have been entered into the licensees corrective action program. The violation and condition report are listed in Section 4OA7.

REPORT DETAILS

Summary of Plant Status

Grand Gulf Nuclear Station began the inspection period at 60 percent rated thermal power due to fuel leak location testing, which began the previous quarter. During the inspection period, the plant was limited to 96 percent power due to the isolation of the second-stage steam to both the A and B moisture separator reheaters on January 9, 2011.

• On June 29, 2011, after locating and suppressing the fuel leak, the plant was returned to 96 percent power.

• On July 9, 2011, operators reduced power to 63 percent for a planned control rod sequence exchange, control rod testing, and turbine testing. The plant was returned to 96 percent power on July 10, 2011.

• On August 5, 2011, operators reduced power to 75 percent for control rod testing, control rod friction testing and turbine testing. The plant was returned to 96 percent power on August 7, 2011.

• On August 12, 2011, operators reduced power to 94.5 percent to remove the heater drain pump B from service to repair a steam leak on a pipe plug on the pump casing.

The plant was returned to 96 percent power the same day.

• On September 1, 2011, operators reduced power to 85 percent for planned control rod testing and turbine testing. The plant was returned to 96 percent power on September 2,

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

1R04 Equipment Alignments

.1 Partial Walkdown

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

• Standby gas treatment system A during a maintenance outage of standby gas treatment system B

• Standby fresh air system B during a surveillance run of standby fresh air system A

• Division 3 emergency diesel generator following a surveillance run The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Updated Final Safety Analysis Report, technical specification requirements, administrative technical specifications, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also inspected accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three partial system walkdown samples as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns that were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

• Division 1 switchgear room (OC202)

• Division 2 switchgear room (OC215)

• Division 1 and 2 remote shutdown rooms and division 3 switch gear room (OC208, OC208A, and OC210)

• Division 1 and 2 reactor protection motor generator set rooms (OC407, OC409, OC707, and OC709)

• Upper and lower cable spreading rooms (OC401, OC410 and OC702)

The inspectors reviewed areas to assess if licensee personnel had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant; effectively maintained fire detection and suppression capability; maintained passive fire protection features in good material condition; and had implemented adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features, in accordance with the licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to affect equipment that could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event. Using the documents listed in the attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed; that transient material loading was within the analyzed limits; and that fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensees corrective action program.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of five quarterly fire-protection inspection samples as defined in Inspection Procedure 71111.05-05.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

.1 Quarterly Review

a. Inspection Scope

On August 3, 2011, the inspectors observed a crew of licensed operators in the plants simulator to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

• Licensed operator performance

• Crews clarity and formality of communications

• Crews ability to take timely actions in the conservative direction

• Crews prioritization, interpretation, and verification of annunciator alarms

• Crews correct use and implementation of abnormal and emergency procedures

• Control board manipulations

• Oversight and direction from supervisors

• Crews ability to identify and implement appropriate technical specification actions and emergency plan actions and notifications The inspectors compared the crews performance in these areas to pre-established operator action expectations and successful critical task completion requirements.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one quarterly licensed-operator requalification program sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

.2 Biennial Inspection

The licensed operator requalification program involves two training cycles that are conducted over a 2-year period. In the first cycle, the annual cycle, the operators are administered an operating test consisting of job performance measures and simulator scenarios. In the second part of the training cycle, the biennial cycle, operators are administered an operating test and a comprehensive written examination.

a. Inspection Scope

To assess the performance effectiveness of the licensed operator requalification program, the inspectors conducted personnel interviews, reviewed both the operating tests and written examinations, and observed ongoing operating test activities.

The inspectors interviewed three licensee personnel, consisting of one operator, one instructor, and one senior operator, to determine their understanding of the policies and practices for administering requalification examinations. The inspectors also reviewed operator performance on the written exams and operating tests. These reviews included observations of portions of the operating tests by the inspectors. The operating tests observed included six job performance measures and two scenarios that were used in the current biennial requalification cycle. These observations allowed the inspectors to assess the licensee's effectiveness in conducting the operating test to ensure operator mastery of the training program content. The inspectors also reviewed medical records of six licensed operators for conformance to license conditions, the licensees system for tracking qualifications, and records of license reactivation for five operators.

The results of these examinations were reviewed to determine the effectiveness of the licensees appraisal of operator performance and to determine if feedback of performance analysis into the requalification training program was being accomplished.

The inspectors interviewed members of the training department and reviewed minutes of

training review group meetings to assess the responsiveness of the licensed operator requalification program to incorporate the lessons learned from both plant and industry events. Examination results were also assessed to determine if they were consistent with the guidance contained in NUREG 1021, "Operator Licensing Examination Standards for Power Reactors", Revision 9, Supplement 1, and NRC Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process."

In addition to the above, the inspectors reviewed examination security measures, simulator fidelity, and existing logs of simulator deficiencies.

On September 6, 2011, the licensee informed the lead inspector of the biennial examination results. The inspector compared these results to Appendix I, Licensed Operator Requalification Significance Determination Process.

The inspectors completed one inspection sample of the biennial licensed operator requalification program.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk significant system:

• Neutron monitoring system (C51)

The inspectors reviewed events such as where ineffective equipment maintenance has resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• Implementing appropriate work practices

• Identifying and addressing common cause failures

• Scoping of systems in accordance with 10 CFR Part 50.65(b)

• Characterizing system reliability issues for performance

• Charging unavailability for performance

• Trending key parameters for condition monitoring

• Ensuring proper classification in accordance with 10 CFR Part 50.65(a)(1) or -

(a)(2)

• Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance through preventive maintenance, as described in 10 CFR Part 50.65(a)(2), or as requiring the establishment of appropriate and adequate goals and corrective actions for systems classified as not having adequate performance, as described in 10 CFR Part 50.65(a)(1)

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

The inspectors also performed a review of the (a)(3) Periodic Evaluation. This review is credited as an inspection sample.

These activities constitute completion of two quarterly maintenance effectiveness samples as defined in Inspection Procedure 71111.12-05.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed licensee personnel's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

• The week of July 11, 2011, while the plant placed the mitigating monitor system in service after a design change

• The week of August 1, 2011, while the plant performed a system outage on the standby gas treatment system B

• The week of August 22, 2011, while the plant performed maintenance outage on the diesel driven fire pump A and the division 1 containment and drywell hydrogen analyzers

• The week of September 19, 2011, during emergent issues with a bearing replacement for reactor protection system motor generator set B, resulting in the licensee having to enter yellow risk The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that licensee personnel performed risk assessments as required by 10 CFR Part 50.65(a)(4) and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four maintenance risk assessments and emergent work control inspection samples as defined in Inspection Procedure 71111.13-05.

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• The interface of the mitigating monitoring system and the plant chilled water system not having a radiation monitor on the plant chilled water system

• Standby service water system B degraded bolts on return flange

• Hydrogen igniter division 1 fuse failure

• Anticipated transient without a scram alternate rod insertion degraded batteries The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that technical specification operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and

design criteria in the appropriate sections of the technical specifications and Update Final Safety Analysis Report to the licensee personnels evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four operability evaluations inspection samples as defined in Inspection Procedure 71111.15-04.

b. Findings

Introduction.

The inspectors reviewed a self-revealing, Green 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.

Description.

On August 4, 2011, the inspectors were performing an operability review of condition report CR-GGN-2011-05388. Night shift operators were attempting to run the division 1 hydrogen igniters and determined that half the division 1 igniters would not energize. The licensee investigated the loss of power to the igniters and determined that one of the fuses for the division 1 hydrogen igniter control circuit had blown. The licensee also determined that the blown fuse was 0.3 amps and should have been 0.8 amps per plant drawings. The shift manager declared the division 1 hydrogen igniters inoperable and entered the 30 day shutdown limiting condition of operation. The shift manager directed the maintenance department to replace the blown fuse with an available 0.3 amp fuse until engineering could perform an evaluation of the circuit while keeping the division 1 hydrogen igniters in an inoperable status. Through a review of site drawings and calculations, it was determined that the correct fuse size for the circuit was 0.8 amps. The licensee performed an operability determination for the 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 replaced the incorrect fuses one division at a time with the correct size 0.8 amp fuses and restored the hydrogen igniters to operable status.

The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2011-05388. The short term corrective action included replacing the fuses with the correct size. Additionally, the licensee conducted a review of their documents to determine when the wrong fuses were installed. The licensee concluded that wrong fuses were installed at plant startup.

Analysis.

The inspectors determined that the failure to install the correct size fuse in the control circuit of the hydrogen igniters is a performance deficiency. This finding is more

than minor because it is associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affects 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 maintained functionality over the life of the plant based on satisfactory surveillance tests and had 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 was 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.

Enforcement.

Title 10 of Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established for the selection and review of suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems, and components.

Contrary to the above, on August 4, 2011, and before, the licensee failed to ensure the correct fuses were in installed in the hydrogen igniter control circuits during the startup of the plant. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-05388. Because this finding was determined to be of very low safety significance and was entered into the licensees corrective action program, this violation is being treated as a noncited violation, consistent with Section 2.3.2a of the NRC Enforcement Policy: NCV 05000416/2011004-01, Failure to Ensure Correct Fuses were Installed in the Hydrogen Igniter Control Circuits.

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors reviewed the following postmaintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• Standby gas treatment system B after scheduled maintenance

• Diesel driven fire pump A after scheduled maintenance

• Standby service water system B fan C after scheduled maintenance

• Containment inner door seal after scheduled maintenance

• Reactor protection system A and reactor protection system B motor generator set after breaker work on A and bearing replacement on B motor generator set

• Safety related switchgear room ventilation fan bearing replacement The inspectors selected these activities based upon the structure, system, or component's ability to affect risk. The inspectors evaluated these activities for the following:

• The effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed

• Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate The inspectors evaluated the activities against the technical specifications, the Update Final Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with postmaintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of six postmaintenance testing inspection samples as defined in Inspection Procedure 71111.19-05.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report, procedure requirements, and technical specifications to ensure that the surveillance activities listed below demonstrated that the systems, structures, and/or components tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the significant surveillance test attributes were adequate to address the following:

• Preconditioning

• Evaluation of testing impact on the plant

• Acceptance criteria

• Test equipment

• Procedures

• Jumper/lifted lead controls

• Test data

• Testing frequency and method demonstrated technical specification operability

• Test equipment removal

• Restoration of plant systems

• Fulfillment of ASME Code requirements

• Updating of performance indicator data

• Engineering evaluations, root causes, and bases for returning tested systems, structures, and components not meeting the test acceptance criteria were correct

• Reference setting data

• Annunciators and alarms set-points The inspectors also verified that licensee personnel identified and implemented any needed corrective actions associated with the surveillance testing.

• On July 5, 2011, division 2, 125 Vdc battery charger 1B4

• On August 3, 2011, residual heat removal A quarterly inservice test

• On August 13, 2011, containment isolation valves 1E61-F009 and 1E61-F010

• On September 16, 2011, division 3 high pressure core spray diesel generator

• On September 20, 2011, main steam line high flow functional test channels 1A, 2A, 3A, 4A, 1C, 2C, 3C, and 4C Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of five surveillance testing inspection samples as defined in Inspection Procedure 71111.22-05.

b. Findings

No findings were identified.

1EP6 Drill Evaluation

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of a routine licensee emergency drill on July 19, 2011, to identify any weaknesses and deficiencies in classification, notification, and protective action recommendation development activities. The inspectors observed emergency response operations in the control room simulator and the emergency operations facility to determine whether the event classification, notifications, and protective action recommendations were performed in accordance with procedures. The inspectors also attended the licensee drill critique to compare any inspector-observed weakness with those identified by the licensee staff in order to evaluate the critique and to verify whether the licensee staff was properly identifying weaknesses and entering them into the corrective action program. As part of the inspection, the inspectors reviewed the drill package and other documents listed in the attachment.

These activities constitute completion of one sample as defined in Inspection Procedure 71114.06-05.

b. Findings

No findings were identified.

2RS0 6 Radioactive Gaseous and Liquid Effluent Treatment

a. Inspection Scope

This area was inspected to:

(1) ensure the gaseous and liquid effluent processing systems are maintained so radiological discharges are properly mitigated, monitored, and evaluated with respect to public exposure;

(2) ensure abnormal radioactive gaseous or liquid discharges and conditions, when effluent radiation monitors are out-of-service, are controlled in accordance with the applicable regulatory requirements and licensee procedures;

(3) verify the licensees quality control program ensures the radioactive effluent sampling and analysis requirements are satisfied so discharges of radioactive materials are adequately quantified and evaluated; and

(4) verify the adequacy of public dose projections resulting from radioactive effluent discharges. The inspectors used the

requirements in 10 CFR Part 20; 10 CFR Part 50, Appendices A and I; 40 CFR Part 190; the Offsite Dose Calculation Manual, and licensee procedures required by the Technical Specifications as criteria for determining compliance. The inspectors interviewed licensee personnel and reviewed and/or observed the following items:

• Radiological effluent release reports since the previous inspection and reports related to the effluent program issued since the previous inspection, if any

• Effluent program implementing procedures, including sampling, monitor setpoint determinations and dose calculations

• Equipment configuration and flow paths of selected gaseous and liquid discharge system components, filtered ventilation system material condition, and significant changes to their effluent release points, if any, and associated 10 CFR Part 50.59 reviews

• Selected portions of the routine processing and discharge of radioactive gaseous and liquid effluents (including sample collection and analysis)

• Controls used to ensure representative sampling and appropriate compensatory sampling

• Results of the inter-laboratory comparison program

• Effluent stack flow rates

• Surveillance test results of technical specification-required ventilation effluent discharge systems since the previous inspection

• Significant changes in reported dose values, if any

• A selection of radioactive liquid and gaseous waste discharge permits

• Part 61 analyses and methods used to determine which isotopes are included in the source term

• Offsite dose calculation manual changes, if any

• Meteorological dispersion and deposition factors

• Latest land use census

• Records of abnormal gaseous or liquid tank discharges, if any

• Groundwater monitoring results

• Changes to the licensees written program for indentifying and controlling contaminated spills/leaks to groundwater, if any

• Identified leakage or spill events and entries made into 10 CFR Part 50.75 (g)records, if any, and associated evaluations of the extent of the contamination and the radiological source term

• Offsite notifications, and reports of events associated with spills, leaks, or groundwater monitoring results, if any

• Audits, self-assessments, reports, and corrective action documents related to radioactive gaseous and liquid effluent treatment since the last inspection Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of the one required sample, as defined in Inspection Procedure 71124.06-05.

b. Findings

No findings were identified.

2RS0 7 Radiological Environmental Monitoring Program

a. Inspection Scope

This area was inspected to:

(1) ensure that the radiological environmental monitoring program verifies the impact of radioactive effluent releases to the environment and sufficiently validates the integrity of the radioactive gaseous and liquid effluent release program;

(2) verify that the radiological environmental monitoring program is implemented consistent with the licensees technical specifications and/or offsite dose calculation manual, and to validate that the radioactive effluent release program meets the design objective contained in Appendix I to 10 CFR Part 50; and

(3) ensure that the radiological environmental monitoring program monitors non-effluent exposure pathways, is based on sound principles and assumptions, and validates that doses to members of the public are within the dose limits of 10 CFR Part 20 and 40 CFR Part 190, as applicable. The inspectors reviewed and/or observed the following items:

• Annual environmental monitoring reports and offsite dose calculation manual

• Selected air sampling and thermoluminescence dosimeter monitoring stations

• Collection and preparation of environmental samples

• Operability, calibration, and maintenance of meteorological instruments

• Selected events documented in the annual environmental monitoring report which involved a missed sample, inoperable sampler, lost thermoluminescence dosimeter, or anomalous measurement

• Selected structures, systems, or components that may contain licensed material and have a credible mechanism for licensed material to reach ground water

• Records required by 10 CFR Part 50.75(g)

• Significant changes made by the licensee to the offsite dose calculation manual as the result of changes to the land census or sampler station modifications since the last inspection

• Calibration and maintenance records for selected air samplers, composite water samplers, and environmental sample radiation measurement instrumentation

• Inter-laboratory comparison program results

• Audits, self-assessments, reports, and corrective action documents related to the radiological environmental monitoring program since the last inspection Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of the one required sample as defined in Inspection Procedure 71124.07-05.

b. Findings

No findings were identified.

2RS08 Radioactive Solid Waste Processing, and Radioactive Material Handling, Storage, and Transportation (71124.08)

a. Inspection Scope

This area was inspected to verify the effectiveness of the licensees programs for processing, handling, storage, and transportation of radioactive material. The inspectors used the requirements of 10 CFR Parts 20, 61, and 71 and Department of Transportation regulations contained in 49 CFR Parts 171-180 for determining compliance. The inspectors interviewed licensee personnel and reviewed the following items:

• The solid radioactive waste system description, process control program, and the scope of the licensees audit program

• Control of radioactive waste storage areas including container labeling/marking and monitoring containers for deformation or signs of waste decomposition

• Changes to the liquid and solid waste processing system configuration including a review of waste processing equipment that is not operational or abandoned in place

• Radio-chemical sample analysis results for radioactive waste streams and use of scaling factors and calculations to account for difficult-to-measure radionuclides

• Processes for waste classification including use of scaling factors and 10 CFR Part 61 analysis

• Shipment packaging, surveying, labeling, marking, placarding, vehicle checking, driver instructing, and preparation of the disposal manifest

• Audits, self-assessments, reports, and corrective action reports radioactive solid waste processing, and radioactive material handling, storage, and transportation performed since the last inspection Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of the one required sample as defined in Inspection Procedure 71124.08-05.

b. Findings

No findings were identified.

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the performance indicator data submitted by the licensee for the second quarter 2011 performance indicators for any obvious inconsistencies prior to its public release in accordance with Inspection Manual Chapter 0608, Performance Indicator Program.

This review was performed as part of the inspectors normal plant status activities and, as such, did not constitute a separate inspection sample.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensees corrective action program at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. The inspectors reviewed attributes that included the complete and accurate identification of the problem; the timely correction, commensurate with the safety significance; the evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrences reviews; and the classification, prioritization, focus, and timeliness of corrective actions. Minor issues entered into the licensees corrective action program because of the inspectors observations are included in the attached list of documents reviewed.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure, they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

No findings were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program. The inspectors accomplished this through review of the stations daily corrective action documents.

The inspectors performed these daily reviews as part of their daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings were identified.

.3 Selected Issue Follow-up Inspection: Inadequate Preventive Maintenance for the

Refueling Platform

a. Inspection Scope

During a review of items entered in the licensees corrective action program, the inspectors recognized a corrective action item documenting a loss of control of the fuel handling bridge during the licensees dry cask storage campaign. The inspectors reviewed the apparent cause evaluation and associated corrective actions.

These activities constitute completion of one in-depth problem identification and resolution sample as defined in Inspection Procedure 71152-05.

b. Findings

Introduction.

The inspectors reviewed a self-revealing, Green 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.

Description.

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 fuel handling platform is electrically controlled by a paddle switch that can be moved in the forward and reverse direction and has a spring return to neutral. When the fuel handling operator releases the paddle, it should return to neutral, which stops the movement of the motor. The bridge then coasts to a neutral state until the brake sets and stops the bridge. The fuel handling bridge was entering the horizontal fuel transfer canal when the operator released the paddle switch 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. Efforts to bring the bridge to a stop by manipulating the paddle switch to the forward direction failed to prevent the bridge from tripping the zone limit switch. When the zone limit switch was actuated, the motor returned to a neutral state, and the bridge stopped as designed.

The zone limit switch is the last defense preventing the fuel handling bridge from colliding into the fuel pool wall, and it performed its function as designed.

The licensee determined that the fuel bridge did not stop when the paddle was released because the paddle switch was stuck in the reverse direction. The licensee was aware that the fuel handling bridge paddle switches were prone to sticking for various reasons.

In 2001, a condition report was written documenting that the grease used in the paddle switches can evaporate and become sticky over time. Furthermore, dust and moisture can affect the lubrication capability of the grease. The licensee concluded that the switches have to be cleaned, adjusted, and re-greased periodically to ensure proper operation. The preventative maintenance strategy the licensee developed was based on an 18-month period to coincide with refueling outages. During the implementation of the dry fuel storage campaign, which occurs between refueling outages, the licensee did not

perform the preventative maintenance on the paddle switch prior to utilizing the fuel handling bridge.

The licensee documented this issue in Condition Report CR-GGN-2011-04896.

Corrective actions included replacing the paddle switch and adjusting the preventive maintenance instructions to include cleaning and re-greasing the paddle switch on the fuel handling platform before every dry fuel cask campaign.

Analysis.

The inspectors determined that the failure to implement preventative maintenance prior to using the fuel handling bridge in support of the dry fuel storage campaign is a performance deficiency. The finding is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity Cornerstone and adversely affects 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. 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 has 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).

Enforcement.

Title 10 CFR Part 50 Appendix B, Criterion V, states, in part, 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 and drawings. Contrary to the above, on and before July 15, 2011, the licensee failed to prescribe preventative maintenance instructions for the fuel handling bridge paddle switch prior to implementing the dry cask loading campaign. The finding was entered into the corrective action program Condition Report CR-GGN-2011-04896. Because the finding was determined to be of very low safety significance and was entered into the licensees corrective action program, this violation is being treated as a noncited violation consistent with Section 2.3.2a of the NRC Enforcement Policy: NCV 05000416/2011004-02, Failure to Perform Preventative Maintenance on the Fuel Handling Bridge Paddle Switch.

.4 Selected Issue Follow-up Inspection: Inadequate Preventive Maintenance for the Critical

Equipment

a. Inspection Scope

During a review of items entered in the licensees corrective action program, the inspectors recognized a corrective action item documenting excessive relay cycling associated with NUS thermal switches. The inspectors reviewed apparent cause evaluations and associated corrective actions.

These activities constitute completion of one in-depth problem identification and resolution sample as defined in Inspection Procedure 71152-05.

b. Findings

Introduction.

The inspectors reviewed a self-revealing, Green 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.

Description.

While reviewing condition reports associated with NUS thermal switches, the inspectors noted the following two examples of failures of safety related equipment that occurred as a result of the licensees failure to perform appropriate preventive maintenance on components:

• 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 immediate corrective action was to replace the contactor coil and to restore the room cooler.

In both cases, the failed components were original plant equipment. The inspectors determined that the failed components were correctly classified as high-critical, but that the licensee had not established the need for any preventive-maintenance measures for these components as required by procedure.. The licensee entered these issues into the corrective action program as Condition Reports CR-GGN-2011-3730 and CR-GGN-2011-4313. Immediate corrective actions implemented by the licensee included establishing preventive maintenance measures for the failed equipment and identifying any additional components that were used in critical systems that had not been previously identified.

Procedure EN-DC-335, Rev. 3, PM Basis Template, requires the licensee to perform preventive-maintenance evaluations for all high-critical components. Procedure EN-DC-153, Preventive Maintenance Component Classification, revision 6, says that any component whose failure causes the unit to enter a 72-hour shutdown Technical Specification action statement is a high-critical component.

Analysis.

The inspectors determined that the licensees failure to follow procedure EN-DC-335 and evaluate the damper relays in the standby service water B pump house ventilation system and contactor coils in breaker 52-154128 for preventative maintenance was a performance deficiency. The finding is more than minor because it is associated 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, so its cause is not indicative of current performance; therefore, no cross-cutting aspect was identified.

Enforcement.

Technical Specification 5.4.1.a requires that written procedures be established, implemented, and maintained as recommended by NRC Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 9a, includes procedures for maintenance that can affect the performance of safety-related equipment. Procedure EN-DC-335, Rev. 3, PM Basis Template, requires a preventive maintenance basis template be applied to all critical components.

Contrary to the above, before June 2, 2011, the licensee failed to apply a preventive maintenance basis template to two critical components, in that:

• damper relays in the standby service water B pump house ventilation system and contactor coils in breaker 52-154128 were high-critical components, and

• the licensee failed to apply preventive maintenance basis template to damper relays in the standby service water B pump house ventilation system and contactor coils in breaker 52-154128.

The finding was entered into the licensees corrective action program as Condition Reports CR-GGN-2011-3730 and CR-GGN-2011-4313. Because this finding was determined to be of very low safety significance and was entered into the licensees corrective action program, this violation is being treated as a noncited violation, consistent with Section 2.3.2a of the NRC Enforcement Policy:

NCV 05000416/2011004-03, Failure to Establish Preventive Maintenance for Components Used in Critical Applications.

.5 In-depth Review of Operator Workarounds

a. Inspection Scope

The inspectors evaluated the licensees implementation of their process used to identify, document, track, and resolve operational challenges. Inspection activities included, but were not limited to, a review of the cumulative effects of the operator workarounds on system availability and the potential for improper operation of the system, for potential impacts on multiple systems, and on the ability of operators to respond to plant transients or accidents.

The inspectors performed a review of the cumulative effects of operator workarounds.

The documents listed in the attachment were reviewed to accomplish the objectives of the inspection procedure. The inspectors reviewed current operational challenge records to determine whether the licensee was identifying operator challenges at an appropriate threshold, had entered them into their corrective action program, and had proposed or implemented appropriate and timely corrective actions, which addressed each issue. Reviews were conducted to determine if any operator challenge could increase the possibility of an initiating event, if the challenge was contrary to training, required a change from long-standing operational practices, or if it created the potential for inappropriate compensatory actions. Additionally, all temporary modifications were reviewed to identify any potential effect on the functionality of mitigating systems, impaired access to equipment, or required equipment uses for which the equipment was not designed. Daily plant and equipment status logs, degraded instrument logs, and operator aids or tools being used to compensate for material deficiencies were also assessed to identify any potential sources of unidentified operator workarounds.

These activities constitute completion of one operator workarounds annual inspection sample as defined in Inspection Procedure 71152.

b. Findings

No findings were identified.

4OA3 Event Follow-up

.1 Reactor Protection System Power Supply Issues

On September 3, 2011, the output breaker on the division 2 reactor protection system motor generator set tripped, causing a division 2 half scram. The operators responded by entering the Loss of RPS Bus, off normal event procedure and transferred the reactor protection system to the alternate power supply, reset the half scram, and exited the off normal event procedure. On September 11, 2011, while the reactor protection system B was still aligned with the alternate power supply, the division 1 reactor protector system motor generator set electrical protection assembly breaker C71S003C tripped, causing a division 1 half scram. The operators declared the electrical protection assembly breaker inoperable and entered an unplanned 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limiting condition of operation. The operators reset the electrical protection assembly breaker and re-energized the motor generator set to supply power to the division 1 reactor protection system and reset the half scram. The operators remained in the limiting condition of operation until the cause of the electrical protection assembly breaker trip could be determined.

The Grand Gulf Nuclear Station Updated Final Safety Analysis Report and reactor protection system operating instruction prohibited placing both divisions of the reactor protection system on the alternate power supply concurrently. The licensee proceeded down three parallel paths to exit the limiting condition of operation, one of which was to change the Updated Final Safety Analysis Report and station procedures via the 10 CFR Part 50.59 process to allow the concurrent use of alternate power supply for both

divisions of the reactor protection systems. On September 13, 2011, the licensee completed the process applicability determination that allowed the change to the Updated Final Safety Analysis Report and plant procedures. Prior to the expiration of the limiting condition of operation, the operators aligned the division 1 reactor protection system to its alternate power supply and exited the limiting condition of operation. The inspectors monitored the licensees activities to exit the unplanned limiting condition of operation, which included attending meetings, evaluating the license basis document change request for changing the Updated Final Safety Analysis Report and plant procedures, and interfacing with licensee management to determine their plan of action.

Specific documents reviewed during this event follow-up are listed in the attachment.

These activities constitute completion of one event follow-up as defined in Inspection Procedure 71153-05.

b. Findings

No findings were identified.

4OA6 Meetings

Exit Meeting Summary

The inspectors debriefed Marty Richey, Director of Nuclear Safety Assurance, and other members of the licensee staff of the results of the licensed operator requalification program inspection on August 18, 2011, and telephonically exited with Michael Bacon, Superintendent, Simulator and Training Support, on September 26, 2011. The licensee representative acknowledged the inspection results. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.

No proprietary information was identified.

On August 19, 2011, the inspectors presented the results of the radiation safety inspections to Marty Richey, Director of Nuclear Safety Assurance, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On October 11, 2011, the inspectors presented the inspection results to Mike Perito, Site Vice President Operations, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section 2.3.2 of the NRC Enforcement Policy for being dispositioned as a noncited violation.

.1 Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings,

states, in part, that activities affecting quality shall be accomplished in accordance with prescribed procedures of a type appropriate to the circumstances. Contrary to this requirement, on July 15, 2011, the refueling bridge operator was not able to stop the fuel handling platform when the paddle switch was released from the reverse position. The fuel handling platform stopped due to the zone limit switch actuating, the motor returned to a neutral state, and the bridge stopped as designed. The zone limit switch is the last defense preventing the fuel handling bridge from colliding into the fuel pool wall, and it performed its function as designed. The licensee stopped all fuel movement and determined that the paddle switch had malfunctioned due to the failure to perform the preventive maintenance tasks of cleaning, adjusting, and re-greasing the paddle switches on the fuel handling platform prior to the start of the dry fuel cask campaign. A contributing cause to this event was a lack of procedure guidance provided in 04-1-01-F11-3, Fuel Handling Platform, Revision 41. The refueling platform procedure did not have any step to direct the operator to immediately depress the stop button if the fuel handling platform is not operating as intended. The switch was replaced, returned to service, and the licensee revised the procedure to include a step to direct the operator to immediately depress the stop button if fuel handling platform was not operating as intended. This issue was entered into the licensees corrective action program as Condition Report CR-GGN-2011-04896. The finding was determined to be 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.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

J. Browning, General Plant Manager

M. Bacon, Superintendent, Simulator and Training Support

R. Bevily, Specialist and Training Coordinator, Chemistry and Radiation Protection

D. Bowers, Supervisor/Coordinator, Maintenance

J. Carey, Manager, Training

D. Coulter, Senior Licensing Specialist

R. Crowe, OSC Manager, I & C Support

H. Farris, Assistant Operations Manager

K. Higgenbotham, Planning and Scheduling Manager

J. Houston, Maintenance Manager

D. Jackson, Sr. HP/Chem Specialist, Chemistry

R. Jackson, Senior Licensing Specialist

D. Jones, Manager, Design Engineering

J. Lassetter, Supervisor, Chemistry

C. Lewis, Manager, Emergency Preparedness

C. Loyd, Supervisor, Engineering

J. Miller, Manager, Operations

L. Patterson, Manager, Program Engineering

C. Perino, Licensing Manager

M. Perito, Site Vice President of Operations

M. Richey, Director, Nuclear Safety Assurance

R. Scarbrough, Specialist and Lead Offsite Liaison, Licensing

J. Seiter, Senior Licensing Specialist

J. Shaw, Manager, System Engineering

D. Wiles, Engineering Director

R. Wilson, Manager, Quality Assurance

T. Trichell, Manager, Radiation Protection

NRC Personnel

R. Smith, Senior Resident Inspector

B. Rice, Resident Inspector

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

Failure to Ensure Correct Fuses were Installed in the Hydrogen

05000416/20110004-01 NCV Igniter Control Circuits (Section 1R15)

Failure to Perform Preventative Maintenance on the Fuel

05000416/20110004-02 NCV Handling Bridge Paddle Switch (Section 4OA2)

Failure to Establish Preventative Maintenance for Components

05000416/20110004-03 NCV Used in Critical Applications (Section 4OA2)

Attachment