IR 05000416/2007002

IR 05000416-07-002, on 01/01/07 Through 03/31/07, Entergy Operations, Inc., Grand Gulf, Integrated Resident & Regional Report, Operability Evaluations, Surveillance Testing & Access Control

ML071210585
Person / Time
Site:	Grand Gulf
Issue date:	05/01/2007
From:	Hay M NRC/RGN-IV/DRP/RPB-C
To:	Brian W Entergy Operations
References
IR-07-002
Download: ML071210585 (51)

Text

May 1, 2007

SUBJECT:

GRAND GULF NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000416/2007002

Dear Mr. Brian:

On March 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Grand Gulf Nuclear Station facility. The enclosed integrated report documents the inspection findings, which were discussed on April 3, 2007, with you and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.

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

This report documents seven NRC identified and self-revealing findings of very low safety significance (Green). All seven of these findings were determined to involve violations of NRC requirements. Additionally, four licensee-identified violations determined to be of very low safety significance are 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 noncited violations (NCVs) consistent with Section VI.A of the NRC Enforcement Policy. If you contest this/these NCVs, 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 DC 20555-0001; with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011-4005; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001; and the NRC Resident Inspector at the Grand Gulf Nuclear Station facility.

-2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Michael C. Hay, Chief Project Branch C Division of Reactor Projects Docket: 50-416 License: NPF-29

Enclosure:

Inspection Report 05000416/2007002 w/Attachment: Supplemental Information

REGION IV==

Docket:

50-416 Licenses:

NPF-29 Report No.:

05000416/2007002 Licensee:

Entergy Operations, Inc.

Facility:

Grand Gulf Nuclear Station Location:

Waterloo Road Port Gibson, Mississippi 39150 Dates:

January 1 through March 31, 2007 Inspectors:

G. Miller, Senior Resident Inspector A. Barrett, Resident Inspector D. Bollock, Project Engineer K. Clayton, Operations Engineer P. Goldberg, P.E., Reactor Inspector G. George, Reactor Inspector W. Sifre, Senior Reactor Inspector L. Ricketson, Senior Health Physics Inspector Approved By:

Michael C. Hay, Chief Project Branch C Division of Reactor Projects

ENCLOSURE-2-

SUMMARY OF FINDINGS

IR05000416/2007002; 1/1/07 - 3/31/07; Grand Gulf Nuclear Station -- Integrated Resident and

Regional Report; Operability Evaluations, Surveillance Testing, Access Control to Radiologically Significant Areas, Identification and Resolution of Problems.

This report covered a 3-month period of inspection by resident inspectors and Regional office inspectors. These inspection activities identified seven Green findings, all of which were noncited violations. The significance of most findings is indicated by their color (Green, White,

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

Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management's review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649,

"Reactor Oversight Process," Revision 3, dated July 2000.

Cornerstone: Mitigating Systems

Green.

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).

Green.

The inspectors identified a Green noncited violation of Technical Specification 5.4.1(a) for the 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 (Section 1R22).

Green.

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 an 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 (Section 4OA2).

Green.

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 (Section 4OA2).

Cornerstone: Occupational Radiation Safety

Green.

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 significance because (1) it was not an ALARA finding, (2) there was no overexposure, (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 (Section 2OS1).

Green.

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 workers 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 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> 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 (Section 2OS1).

Green.

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 (Section 2OS1).

Licensee-Identified Violations

Violations of very low safety significance which were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. These violations and corrective actions are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Grand Gulf Nuclear Station began the inspection period at 100 percent rated thermal power.

Except for planned control rod pattern adjustments and control rod drive maintenance and testing, the plant remained at or near full rated thermal power until February 20, 2007, when the plant began coasting down in power before shutting down on March 18, 2007, for scheduled Refueling Outage 15. The plant remained shutdown for the balance of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

.1 Readiness For Seasonal Susceptibilities

a. Inspection Scope

During the week of January 15, 2007, the inspectors completed a review of the licensee's readiness for seasonal susceptibilities involving extreme low temperatures.

The inspectors:

(1) reviewed plant procedures, the Updated Final Safety Analysis Report (UFSAR), and Technical Specifications (TS) to ensure that operator actions defined in adverse weather procedures maintained the readiness of essential systems;

(2) walked down portions of the three systems listed below to ensure that adverse weather protection features were sufficient to support operability, including the ability to perform safe shutdown functions;

(3) evaluated operator staffing levels to ensure the licensee could maintain the readiness of essential systems required by plant procedures; and

(4) reviewed the corrective action program (CAP) to determine if the licensee identified and corrected problems related to adverse weather conditions.

C January 18-19, 2007, fire protection system, emergency diesel generators (EDGs), standby service water system Documents reviewed by the inspectors included:

C Procedure 04-1-03-A30-1, Cold Weather Protection, Revision 18 C

Condition Report CR-GGN-2006-1518 C

Condition Report CR-GGN-2006-0131 C

Modification ER-GG-2003-0121 The inspectors completed one sample.

b. Findings

No findings of significance were identified.

ENCLOSURE

1R04 Equipment Alignment

.1 Partial System Walkdowns

a. Inspection Scope

The inspectors:

(1) walked down portions of the three listed risk important systems and reviewed plant procedures and documents to verify that critical portions of the selected systems were correctly aligned; and

(2) compared deficiencies identified during the walkdown to the licensee's UFSAR and CAP to ensure problems were being identified and corrected.

C January 23, 2007, the inspectors walked down the Division I EDG while the Division II EDG was out of service for planned maintenance.

C January 30, 2007, the inspectors walked down Train B of the standby service water system while Train A was out of service for a planned system outage.

C March 20-21, 2007, the inspectors walked down the alternate decay heat removal system while Train B of shutdown cooling was out of service for planned maintenance.

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed three samples.

b. Findings

No findings of significance were identified.

.2 Complete System Walkdown

a. Inspection Scope

The inspectors:

(1) reviewed plant procedures, drawings, the UFSAR, TSs, and vendor manuals to determine the correct alignment of the 125 volt station battery system;

(2) reviewed outstanding design issues, operator workarounds, and UFSAR documents to determine if open issues affected the functionality of the station battery system; and

(3) verified that the licensee was identifying and resolving equipment alignment problems. Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

ENCLOSURE

1R05 Fire Protection

a. Inspection Scope

Quarterly Inspection The inspectors walked down the six listed plant areas to assess the material condition of active and passive fire protection features and their operational lineup and readiness.

The inspectors:

(1) verified that transient combustibles and hot work activities were controlled in accordance with plant procedures;

(2) observed the condition of fire detection devices to verify they remained functional;

(3) observed fire suppression systems to verify they remained functional and that access to manual actuators was unobstructed;

(4) verified that fire extinguishers and hose stations were provided at their designated locations and that they were in a satisfactory condition;

(5) verified that passive fire protection features (electrical raceway barriers, fire doors, fire dampers, steel fire proofing, penetration seals, and oil collection systems) were in a satisfactory material condition;

(6) verified that adequate compensatory measures were established for degraded or inoperable fire protection features and that the compensatory measures were commensurate with the significance of the deficiency; and

(7) reviewed the UFSAR to determine if the licensee identified and corrected fire protection problems.

C Division I EDG room (Room 1D302)

C Division I switchgear room (Room OC202)

C Division II switchgear room (Room OC215)

C Division III switchgear room (Room OC210)

C Remote shutdown panel room (Room OC208A/B)

C Division II battery room (Room OC211)

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed six samples.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

.1 Annual External Flooding

a. Inspection Scope

The inspectors:

(1) reviewed the UFSAR, the flooding analysis, and plant procedures to assess seasonal susceptibilities involving external flooding;

(2) reviewed the UFSAR and CAP to determine if the licensee identified and corrected flooding problems;

(3) inspected underground bunkers/manholes to verify the adequacy of

(a) sump pumps,

(b) level alarm circuits,

(c) cable splices subject to submergence, and

(d) drainage for bunkers/manholes;

(4) verified that operator actions for coping with

ENCLOSURE flooding can reasonably achieve the desired outcomes; and

(5) walked down the three below listed areas to verify the adequacy of:

(a) equipment seals located below the floodline,

(b) floor and wall penetration seals,

(c) watertight door seals,

(d) common drain lines and sumps,

(e) sump pumps, level alarms, and control circuits, and

(f) temporary or removable flood barriers.

C February 28 - March 1, 2007, diesel generator building, standby service water pump houses, Culvert 1 Documents reviewed by the inspectors included:

C Calculation CC-Q1Y23-91032, PMP Evaluation for Phase I Road and Yard Paving, Revision 1 C

Calculation CC-Q1Y23-91047, PMP Site Drainage, Revision 0 C

Condition Report CR-GGN-2006-4149 The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection Activities

.1 Inspection Activities Other Than Steam Generator Tube Inspections, Pressurized Water

Reactor Vessel Upper Head Penetration Inspections, Boric Acid Corrosion Control

a. Inspection Scope

The scope of this inspection is to verify that inservice inspection activities are being performed in accordance with American Society of Mechanical Engineers (ASME) Code and other applicable regulatory requirements. The scope of this inspection is to review activities associated with reactor coolant system pressure boundaries and piping connected to the reactor coolant system, reactor vessel internals, and other risk significant piping system boundaries. The inspectors focused the inspection by selecting a majority of components from the reactor recirculation system and steam dryer.

The inspectors reviewed three ultrasonic examinations and five surface examinations.

From those eight examinations, the inspectors observed two ultrasonic examinations. In addition, the inspectors observed one visual examination of the steam dryer. The inspectors verified that each examiner held qualifications to perform each examination.

ENCLOSURE Partial List of Records Report No.

Component Component ID Method ISI-MT-07-001 Reactor Pressure Vessel Circumferential Vessel Head-to-Flange 1B13-AG MT ISI-UT-07-008 Reactor Recirculation System Circ Pipe to Valve B33F023A 1B33G001W4 UT ISI-UT-07-007 Reactor Water Cleanup System Circ Flued Head-to-Pipe 1G33G002W19 UT ISI-UT-07-009 Reactor Recirculation System Circ Pipe-to-Pipe End Cap 1B33G10-B1-C UT ISI-PT-07-002 Reactor Recirculation System Weld 1B33C001B-B4LUG4 PT ISI-PT-07-001 Reactor Recirculation System Weld 1B33C001B-B5LUG5 PT ISI-PT-07-003 Reactor Recirculation System Weld 1B33C001B-B3LUG1 PT ISI-PT-07-004 Reactor Recirculation System Weld 1B33C001B-B6LUG6 PT The inspectors reviewed the site procedures to verify that recordable indications were dispositioned in accordance with ASME Code or an NRC approved alternative. During the performance of the inspection activities, no recordable indications were identified or accepted for continued service.

The inspection procedure requires verification of one to three welds that the welding process and welding examinations were performed in accordance with ASME Code Class 1 or 2 requirements or an NRC approved alternative. No welding was performed on Class 1 or 2 systems during the inspection.

The inspectors completed the one sample required for boiling water reactors per Inspection Procedure 71111.08.

b. Findings

No findings of significance were identified.

ENCLOSURE

.2 Identification and Resolution of Problems

a. Inspection Scope

The inspectors reviewed eight condition reports which dealt with inservice inspection activities, and found that the corrective actions were appropriate. From this review, the inspectors concluded that the licensee had an appropriate threshold for entering issues into the corrective action program and has procedures that direct a root cause evaluation when necessary. The licensee also had an effective program for applying industry operating experience.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

.1 Quarterly Inspection

a. Inspection Scope

The inspectors observed testing and training of senior reactor operators and reactor operators to assess training, operator performance, and the evaluator's critique. The training scenario, GSMS-LOR-00178, Revision 01, involved a loss of the Division 2 safety-related bus, control rod drift, and an anticipated transient without scram.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.2 Biennial Inspection

a. Inspection Scope

The following inspection activities were performed using Inspection Procedure 71111.11, "Licensed Operator Requalification Program," and 10 CFR Part 55.46, "Simulation Facilities," as acceptance criteria. The purpose of this review was to determine if the simulator was capable of supporting initial examinations, supporting requalification training required for all licensed-operators on shift, and supporting reactivity and control manipulations for initial license applications. The licensee communicated to the inspector that they had never used the simulator for reactivity manipulation credits on initial applications and indicated that they did not intend to use the simulator for these reactivity credits on the March 2007 initial exam.

The inspector reviewed the simulator annual performance test book for 2006, in which all annual tests were conducted between September and December of 2006. For

ENCLOSURE simulator testing, the licensee is committed to ANS/ANSI 3.5 -1998, "Nuclear Power Plant Simulators for Use in Operator Training and Examination," in Procedure EN-TQ-202, Revision 3, entitled "Simulator Configuration Control." A sample of core performance test documents were reviewed in order to assess the adequacy of the simulator in supporting reactivity and control manipulations for future exams.

Two transient tests, one scenario, and a work package closeout test were run on the simulator with data capture enabled to verify data collected from previous tests was an accurate representation of the test data run during the testing in October 2006, and also as a verification of reasonable model performance based on the current design of the plant. These tests were:

(1) Dual Feed Pump Trip-Transient Test Two; and

(2) Design Basis Loss of Coolant Accident with Subsequent Loss of Off-Site Power-Transient Test Eight. The analyzed scenario was a Station Blackout event. The work modification package for the Low Pressure Feedwater Heater string inlet and outlet valves (installed in the plant and the simulator) was also run on the simulator.

As part of this review, the inspector interviewed one instructor, one evaluator, two reactor operators, two senior reactor operators, both simulator engineers, and the simulator support supervisor. The interviews were performed in order to collect feedback regarding the fidelity of the simulator, the simulator discrepancy reporting system effectiveness, and training on differences between the simulator and the plant.

The licensee communicated to the inspector that an upgrade request from the simulator staff to senior management for the thermal-hydraulic model was initiated in 2002 and has been discussed for several years but has not been approved. The licensee communicated to the inspector that: 1) the input-output hardware is scheduled for replacement in 2008 and 2) the thermal hydraulic model upgrade should be placed on the two year budgeting plan such that the upgrade would be completed in late 2008 or early 2009.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

.1 Routine Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the following two maintenance rule scoped systems that have displayed performance problems to:

(1) verify the appropriate handling of structure, system, and component (SSC) performance or condition problems;

(2) verify the appropriate handling of degraded SSC functional performance;

(3) evaluate the role of work practices and common cause problems; and

(4) evaluate the handling of SSC issues reviewed under the requirements of the maintenance rule, 10 CFR Part 50 Appendix B, and the TS.

• Suppression pool makeup system level instrumentation (E30)

ENCLOSURE

• Auxiliary building cranes and hoists (T31)

Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed two samples.

b. Findings

No findings of significance were identified.

.2 Triennial Periodic Evaluation

a. Inspection Scope

The inspector reviewed the Grand Gulf Nuclear Station report documenting the last periodic evaluation in accordance with 10 CFR 50.65(a)(3), which was the licensees Maintenance Rule Periodic Assessment for the period from January 1, 2004, through October 18, 2005.

The inspector reviewed the monitoring of risk significant structures, systems, and components with degraded performance to access the effectiveness of the licensees evaluations and the resulting corrective actions. The performance monitoring of non-risk-significant functions using plant level criteria was also reviewed.

The inspector evaluated whether the report contained adequate assessment of the performance of the Maintenance Rule program as well as conformance with applicable programmatic and regulatory requirements. To accomplish this, the inspector verified that the licensee appropriately and correctly addressed the following attributes in the assessment report:

• Program treatment of non-risk-significant structure, system, and component functions monitored against plant level performance criteria

• Program adjustments made in response to unbalanced reliability and availability

• Application of industry operating experience

• Performance review of Category (a)(1) systems

• Evaluation of the bases for system category status change (e.g., (a)(1) to (a)(2)or (a)(2) to (a)(1))

• Effectiveness of performance and condition monitoring at component, train, system and plant levels

• Review and adjustment of definitions of functional failures Inspection Procedure 71111.12 Triennial, Maintenance Effectiveness, requires a

ENCLOSURE minimum sample of four structures, systems, and components. The inspector reviewed three high risk systems and one structure. The inspection sample consisted of the following:

• Standby Service Water System

• Emergency Diesel Generators

• Instrument Air System

• Cask Crane The inspector reviewed the:

(1) evaluations of the balance of reliability and unavailability for maintenance rule functions,

(2) consideration of industry operating experience,

(3) assessment and management of risk related maintenance activities, and

(4) use of insights from the probabilistic risk assessment to support the maintenance rule program.

While reviewing the cask crane, the inspectors noted that the crane was considered safety related and in the maintenance rule program. However, the only time the crane would be used was during normal operation. The inspectors reviewed the maintenance performed on the crane just prior to the first cask lift and found it adequate and determined the crane was single failure proof. The first time the crane was used for a cask lift was the fall of 2006. The inspectors did not review the maintenance history of the crane from the time it was installed.

The inspectors completed four samples.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

.1 Risk Assessment and Management of Risk

a. Inspection Scope

The inspectors reviewed the four listed assessment activities to verify:

(1) performance of risk assessments when required by 10 CFR 50.65 (a)(4) and licensee procedures prior to changes in plant configuration for maintenance activities and plant operations;

(2) the accuracy, adequacy, and completeness of the information considered in the risk assessment;

(3) that the licensee recognized, and/or entered as applicable, the appropriate licensee-established risk category according to the risk assessment results and licensee procedures; and

(4) that the licensee-identified and corrected problems related to maintenance risk assessments.

• WO 100381, standby service water discharge valve maintenance

• WO 102988, Division 1 EDG outage

• WO 104064, residual heat removal Train C closed loop leakage test

• WO 96203, switchyard breaker testing

ENCLOSURE Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed four samples.

b. Findings

No findings of significance were identified.

.2 Emergent Work Control

a. Inspection Scope

For the two work activities listed below, the inspectors:

(1) verified that the licensee performed actions to minimize the probability of initiating events and maintained the functional capability of mitigating systems and barrier integrity systems;

(2) verified that emergent work-related activities such as troubleshooting, work planning/scheduling, establishing plant conditions, aligning equipment, tagging, temporary modifications, and equipment restoration did not place the plant in an unacceptable configuration; and

(3) reviewed the UFSAR to determine if the licensee identified and corrected risk assessment and emergent work control problems.

• WO 102991, plant chiller Train B freon leak

• WO 51024268, drywell purge compressor maintenance Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed two samples.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors:

(1) reviewed plants status documents such as operator shift logs, emergent work documentation, deferred modifications, and standing orders to determine if an operability evaluation was warranted for degraded components;

(2) referred to the UFSAR and design basis documents to review the technical adequacy of licensee operability evaluations;

(3) evaluated compensatory measures associated with operability evaluations;

(4) determined degraded component impact on any TS;

(5) used the Significance Determination Process to evaluate the risk significance of degraded or inoperable equipment; and

(6) verified that the licensee has identified and implemented appropriate corrective actions associated with degraded components.

• CR-GGN-2007-0262, incorrect containment ventilation rad monitor setpoints

ENCLOSURE

• CR-GGN-2007-0378, Division 1 EDG failure to run

• CR-GGN-2007-0660, nonconforming Division 1 EDG thermostatic control valve

• CR-GGN-2007-0831, increased unidentified leakage in the drywell

• CR-GGN-2007-0927, residual heat removal Train C leakage

• CR-GGN-2007-0174, higher than expected Division 2 battery intercell resistance measurements Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed six samples.

b. Findings

Introduction.

The inspectors identified a Green noncited violation (NCV) of 10 CFR Part 50 Appendix B, Criterion V for a failure to follow procedures which resulted in an inadequate operability evaluation.

Description.

The safeguards switchgear and battery room ventilation system uses outside air to provide cooling to the safety related switchgear and battery rooms in the control building. The system is designed to provide adequate cooling for post-design basis accident heat loads with an outside ambient air temperature of 95 degrees Fahrenheit. On November 12, 2006, an operator and an electrician performing a surveillance test inadvertently shut the system fire dampers, isolating ventilation to the Division 1 and Division 3 switchgear and battery rooms and the remote shutdown panel room. This event is described further in Section 4OA2 of this report.

Following isolation of the ventilation system, control room operators performed an immediate operability determination and declared the affected switchgear operable based on engineering judgement given the low outside air temperatures at that time of year, the observed temperature trend, and the ability to provide additional air flow by blocking open fire doors if needed. The operators issued a corrective action as part of associated Condition Report CR-GGN-2006-4394 for licensee engineers to provide a written evaluation of the room cooling for verification of the immediate operability determination. Licensee engineers completed the evaluation and operators declared the switchgear operable on November 13, 2006. That same day, maintenance technicians completed repairing and reopening the affected ventilation system dampers.

The inspectors concluded no actual loss of operability occurred for the affected switchgear given the low outside air temperatures and the room temperatures during the duration of the event, as well as the available recovery methods, such as manually opening the shut dampers or the use of portable air blowers.

The inspectors reviewed the operability evaluation prepared by the licensee engineering staff and compared it to Calculation MC-Q1Z77-92001, Safeguards Switchgear and Battery Room Ventilation and Cooling Requirement, Revision 3. The inspectors identified several errors in the evaluation, all of which were nonconservative. For example, the switchgear heat load in the operability evaluation contained a transcription error from the source calculation, resulting in the evaluation of a lower post-accident heat load than assumed in the design basis calculation. The evaluation also utilized an

ENCLOSURE incorrect and nonconservative correction factor for the density of the outside air. The evaluation also omitted a term used in the design basis calculation to account for the heat added by the ventilation system fans and motors, which are located in the flowstream of the cooling air to the switchgear rooms.

In addition to the discrepancies listed above, the inspectors also noted the operability evaluation assumed that air at outside ambient temperatures was supplied to each of the affected rooms. The inspectors considered this an invalid and nonconservative assumption given the ventilation configuration. Specifically, air entered the Division 2 switchgear room, then passed through the Division 3 switchgear and battery rooms and the remote shutdown panel room before reaching the Division 1 rooms. By the time the air entered the Division 1 switchgear room, it would have already been heated by the equipment in the other rooms and would no longer be at outside ambient temperature.

The inspectors concluded that the evaluation methodology used was therefore too simplistic to provide an accurate analysis.

The inspectors also noted that the calculations in the operability evaluation were based on the forecast high temperature for November 13, 2006. The inspectors concluded this was a nonconservative assumption since it did not account for the inherent inaccuracies in weather forecasting, nor did it account for the mission time of the affected equipment post-accident. The inspectors also concluded that had the rest of the calculation been adequate, this assumption would have imposed a restriction on the operability evaluation that was not specifically identified to operators as required by Procedure EN-OP-104, Operability Determinations, Revision 2.

Analysis.

The failure to implement station procedures is a performance deficiency. 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.

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.

Enforcement.

Criterion V, Instructions, Procedures and Drawings, of Appendix B to 10 CFR Part 50 states, in part, that activities affecting quality shall be prescribed by documented instructions and shall be accomplished in accordance with those instructions. Contrary to the above, on November 13, 2006, licensee engineers failed to implement section 5.4[2] of EN-OP-104, Operability Determinations, Revision 2, which required operability evaluations to identify any associated restrictions or limitations per Step 6 of Attachment 9.5 of the procedure. Because this violation was of very low safety significance and was entered in the corrective action program as CR-GGN-2007-0554, this violation is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000416/2007002-01, Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation.

ENCLOSURE

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors selected the six listed postmaintenance test activities of risk significant systems or components. For each item, the inspectors:

(1) reviewed the applicable licensing basis and/or design-basis documents to determine the safety functions;

(2) evaluated the safety functions that may have been affected by the maintenance activity; and

(3) reviewed the test procedure to ensure it adequately tested the safety function that may have been affected. The inspectors either witnessed or reviewed test data to verify that acceptance criteria were met, plant impacts were evaluated, test equipment was calibrated, procedures were followed, jumpers were properly controlled, test data results were complete and accurate, test equipment was removed, the system was properly re-aligned, and deficiencies during testing were documented. The inspectors also reviewed the UFSAR to determine if the licensee identified and corrected problems related to post-maintenance testing.

• WO 82338, secondary containment isolation Valve G46F253 rebuild

• WO 99584, remove and replace Division 2 EDG fuel injectors

• WO 81761, Division 1 EDG thermostatic control valve thermal element replacement

• WO 103528, low pressure core spray closed loop leakage testing

• WO 51031312, Penetration 1E30P116 local leak rate test

• WO 106106, Division 1 EDG vibration switch replacement Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed six samples.

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities

a. Inspection Scope

The inspectors reviewed the following risk significant refueling items or outage activities to verify defense in depth commensurate with the outage risk control plan and compliance with the TS:

(1) the risk control plan;

(2) tagging/clearance activities; (3)reactor coolant system instrumentation;

(4) electrical power;

(5) decay heat removal; (6)spent fuel pool cooling;

(7) inventory control;

(8) reactivity control;

(9) drywell initial entry;

(10) reduced inventory conditions;

(11) refueling activities;

(12) cooldown activities;

(12) ENCLOSURE control of heavy loads per Operating Experience Smart Sample FY2007-03; and (13)licensee identification and implementation of appropriate corrective actions associated with refueling and outage activities. The inspectors' drywell inspections included observations of the drywell floor for debris; and of supports, braces, and snubbers for evidence of excessive stress, water hammer, or aging. Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the UFSAR, procedure requirements, and TSs to ensure that the seven listed surveillance activities demonstrated that the SSCs tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the following significant surveillance test attributes were adequate:

(1) preconditioning;

(2) evaluation of testing impact on the plant; (3)acceptance criteria;

(4) test equipment;

(5) procedures;

(6) jumper/lifted lead controls;

(7) test data;

(8) testing frequency and method demonstrated TS operability;

(9) test equipment removal;

(10) restoration of plant systems;

(11) fulfillment of ASME Code requirements;

(12) updating of performance indicator (PI) data;

(13) engineering evaluations, root causes, and bases for returning tested SSCs not meeting the test acceptance criteria were correct;

(14) reference setting data; and

(15) annunciator and alarm setpoints. The inspectors also verified that the licensee identified and implemented any needed corrective actions associated with the surveillance testing.

• January 2, 2007, Division 1 EDG monthly surveillance test per Procedure 06-OP-1P75-M-001, Standby Diesel Generator Functional Test, Revision 127

• January 3-4, 2007, standby service water Train A quarterly inservice test per Procedure 06-OP-1P41-Q-0004, Standby Service Water Loop A Valve and Pump Operability Test, Revision 116

• January 18, 2007, Division 1 load shedding and sequencer monthly surveillance test per Procedure 06-OP-1R21-M-0002, Division 1 and 2 Load Shed Sequencer Functional Test, Revision 101

• January 31, 2007, Division 2 EDG quick start surveillance test per Procedure 06-OP-1P75-V-0003, Standby Diesel Generator Operability Verification, Revision 106

• February 21, 2007, daily calculation of reactor coolant system leakage per 06-OP-1000-D-0001, NAME, Revision 121

ENCLOSURE

• March 15, 2007, jetpump operability daily surveillance test per Procedure 06-RE-1B33-D0001, Jetpump Functional Test, Revision 108

• March 12, 2007, reactor core isolation cooling system closed loop leakage test per Procedure 06-ME-1M61-V-003, Local Leak Rate Test, Low Pressure Water, Revision 103 Documents reviewed by the inspectors are listed in the attachment.

The inspectors completed seven samples.

b. Findings

Introduction.

The inspectors identified a Green NCV of TS 5.4.1(a) for failure to meet procedural requirements for 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.

Description.

On March 12, 2007, the inspector was reviewing a local leak rate surveillance test for the suppression pool level instrumentation which caused both channels of suppression pool level to be non-functional. TSs required swapping the suction for reactor core isolation cooling (RCIC) from the condensate storage tank (CST) to maintain the system operable. The inspector noted that the suction was aligned to the CST and questioned the shift technical advisor (STA) on the RCIC system TS requirements. The STA responded that since RCIC had been declared inoperable, no suction swap was required.

While reviewing control room panels several minutes later, the inspector heard the STA direct a control room operator to realign the RCIC suction to the suppression pool. The inspector noted that after completing the valve manipulations, the operator then asked the STA for the reason for the suction swap. The STA responded that the realignment was per the TSs actions for the inoperability of the suppression pool level instrumentation. The inspector noted that although present in the control room, the control room supervisor (CRS) was not informed of the valve manipulations.

The inspector asked the control room supervisor why the RCIC system had been declared inoperable. The CRS responded that it was for a closed loop leakage test of the system. The inspector then questioned both the STA and the CRS about whether the suction realignment would impact the closed loop leakage test. The CRS said that the RCIC suction was required to be aligned to the CST for the test, and he was not aware of the valve manipulation that had been directed by the STA. The CRS then directed the control room operator to restore the RCIC suction to the CST. Following conversations with the shift manager and operations manager and after prompting by the inspector, the licensee entered this issue into their corrective action program as CR-GGN-2007-1060.

ENCLOSURE

Analysis.

The failure to inform the control room supervisor of plant valve manipulations is a performance deficiency. 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 the loss of operability of the RCIC system. This finding has a crosscutting aspect in the area of human performance associated with work practices because the failure to communicate the realignment of the RCIC system to the control room supervisor prevented the control room supervisor from maintaining proper supervisory oversight of work activities.

Enforcement.

Technical Specification 5.4.1(a) requires written procedures to be implemented as recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Section 4[8](b) of EN-OP-115, Conduct of Operations, requires the Control Room Supervisor to direct the activities of the control room operators to maintain stable plant conditions during all plant evolutions. Contrary to the above, the shift technical advisor directed a reactor operator to manipulate valves without the control room supervisors knowledge. Because this violation was of very low safety significance and was entered into the licensee's corrective action program as CR-GGN-2007-1060, this violation is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000416/2007002-02, Ineffective Command and Control Results in Inappropriate Valve Manipulations.

1EP6 Drill Evaluation

a. Inspection Scope

For the below listed drill contributing to Drill/Exercise Performance and emergency response organization PIs, the inspectors:

(1) observed the training evolution to assess classification, notification, and Protective Action Requirement development activities;

(2) compared identified weaknesses and deficiencies against licensee identified findings to determine whether the licensee is properly identifying failures; and

(3) determined whether licensee performance is in accordance with the guidance of the Nuclear Energy Institute (NEI) 99-02, "Voluntary Submission of Performance Indicator Data,"

acceptance criteria.

• February 21, 2007, the inspectors observed the licensees emergency response organization in the simulator, the Emergency Response Facility, the Technical Support Center, and the Operations Support Center respond to a simulated fire and anticipated transient without scram that led to fuel damage and a release to the atmosphere.

Documents reviewed by the inspectors included:

• GGNS 2007 1st Quarter Emergency Preparedness Drill Evaluators Notebook

ENCLOSURE

• Drill Emergency Notification Forms The inspectors completed one sample.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety [OS]

2OS1 Access Control To Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess the licensees performance in implementing physical and administrative controls for airborne radioactivity areas, radiation areas, high radiation areas, and worker adherence to these controls. The inspector used the requirements in 10 CFR Part 20, the TSs, and the licensees procedures required by TSs as criteria for determining compliance. During the inspection, the inspector interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspector performed independent radiation dose rate measurements and reviewed the following items:

• Performance indicator events and associated documentation packages reported by the licensee in the Occupational Radiation Safety Cornerstone

• Controls (surveys, posting, and barricades) of high radiation, or airborne radioactivity areas

• Radiation work permits, procedures, engineering controls, and air sampler locations

• Conformity of electronic personal dosimeter alarm set points with survey indications and plant policy; workers knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarms

• Barrier integrity and performance of engineering controls in airborne radioactivity areas

• Physical and programmatic controls for highly activated or contaminated materials (non-fuel) stored within spent fuel and other storage pools.

• Self-assessments, audits, licensee event reports, and special reports related to the access control program since the last inspection

• Corrective action documents related to access controls

ENCLOSURE

• Licensee actions in cases of repetitive deficiencies or significant individual deficiencies

• Radiation work permit briefings and worker instructions

• Adequacy of radiological controls, such as required surveys, radiation protection job coverage, and contamination control during job performance

• Dosimetry placement in high radiation work areas with significant dose rate gradients

• Changes in licensee procedural controls of high dose rate - high radiation areas and very high radiation areas

• Controls for special areas that have the potential to become very high radiation areas during certain plant operations

• Posting and locking of entrances to all accessible high dose rate - high radiation areas and very high radiation areas

• Radiation worker and radiation protection technician performance with respect to radiation protection work requirements The inspector completed 20 of the required 21 samples.

.1 Failure to Follow Procedural Guidance and Radiation Work Instructions While Supporting

Radiography Operations

Introduction.

The inspector reviewed a self-revealing, NCV of TS 5.4.1 resulting from a failure to follow procedural guidance and radiation work permit instructions while supporting radiography operations. The violation had very low safety significance.

Description.

On March 28, 2007, the licensee conducted radiography on the 161-foot elevation of the containment building. The radiography was scheduled during the morning shift change when fewer workers were present in the area. The radiographer was licensed by the state of Mississippi and the governing procedure for radiation protection personnel was Procedure EN-RP-150, Radiography and X-Ray Testing, Revision 0. The responsible radiation protection supervisor conducted a pre-job briefing, and radiation protection personnel went to their respective assigned locations to barricade and post entrances to the area in which radiography was to be performed.

Some radiation protection technicians assigned to the 208-foot elevation expressed concerns that there were not enough resources to guard each entrance on that elevation and make dose readings during radiography. However, before this issue could be resolved and before each entrance was barricaded and posted, an announcement on the plant paging system stated that radiography operations were commencing. An attempt was made by other radiation protection supervisors to page or call the radiographer and appropriate radiation protection personnel to inform them to halt radiography, but contact

ENCLOSURE was made too late, and the radiographic exposure was completed without confirmation that all entrances were barricaded and posted. Consequently, some entrances were not controlled as required.

All additional radiography was canceled, and radiation protection personnel conducted a search and determined that workers had not entered the radiography area through unbarricaded and unposted entrances. The licensee interviewed the radiographers and the radiation protection personnel that supported radiography and determined the high radiation area on the 161-foot elevation of the containment building was barricaded, posted, and guarded. Additionally, radiation protection personnel determined that no one had received an electronic dosimeter alarm during the time of radiography. Based on this information, the inspector concluded no one had entered the area or received unplanned dose. The licensee documented this occurrence in the corrective action program, initiated fact-finding with the help of the Arkansas Nuclear Ones radiation protection manager, and subsequently concluded that a root cause analysis was necessary before long-term corrective actions were developed.

The licensee had not completed the root cause analysis by the end of the inspection.

However, licensee representatives stated they had determined the responsible radiation protection supervisor had not walked down the area with the radiographer, but had relied upon the radiographer to ensure no unauthorized personnel were in the radiography area and boundaries were barricaded and posted. The inspector noted that the pre-job planning documentation (EN-RP-150, Attachment 9.2) did not require the use of radios.

However, the licensees Procedure EN-RP-150, Section 5.3 [12], stated, If several radiological boundaries have been established where radiographer and radiation protection personnel will perform monitoring, then provide for communication during radiography testing such as radios.

Analysis.

The failure to barricade and post the entrances to the area in which radiography was conducted was a performance deficiency. 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. The occurrence involved the potential for unplanned, unintended dose resulting from actions contrary to licensee procedures and a radiation work permit which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances; therefore, the finding was evaluated using the Occupational Radiation Safety Significance Determination Process. The finding was determined to be of 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 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.

The finding was self-revealing because, when the announcement was made that radiography was commencing, the lack of barricading and posting was readily apparent and the problem was not discovered through a licensee program or process.

ENCLOSURE

Enforcement.

Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Appendix A recommends, in Section 7, procedures for the control of radioactivity for limiting personnel exposure. Procedures EN-RP-150, Radiography and X-Ray Testing, Revision 0, and EN-RP-100, Radworker Expectations, Revision 0, implement this requirement.

Procedure EN-RP-150, Radiography and X-Ray Testing, Revision 0, Section 5.3, Step [8] states, RPT [radiation protection technician] SHALL complete the section titled, Required Steps Prior to Commencing Radiography, of Attachment 9.1, Radiography Testing Pre-Job Brief and Checklist, prior to commencing radiography. Attachment 9.1, Step 12, states, Establish and Verify radiography boundaries. Step 9, states, The radiography area shall be barricaded and posted at 2 millirem per hour with signs stating, Caution Radiation Area, Radiography in Progress, Keep Out. Step 14, states, Prior to the start of radiography, the radiographer AND radiation protection SHALL ensure a walk down of all areas with the radiography boundary has been conducted and ensure unauthorized personnel are cleared from ALL areas enclosed by these barriers.

Procedure RP-100, Radiation Worker Expectations, Revision 0, Section 4, states that it is the responsibility of all individuals to know and follow radiation work permit instructions when performing radiological work. Radiation Work Permit 07-1052, Radiation Protection Instruction 5 states, RP to post Caution Radiation Areas, Radiography in progress, Keep Out boundary at 2 millirem per hour. Radiation Protection Instruction 8 requires, RP and radiographers are to walk down the area and ensure no unauthorized personnel are present.

The licensee violated these requirements when radiation protection personnel working in accordance with Radiation Work Permit 07-1052 did not establish and verify all radiography boundaries. Consequently, all entrances to the area in which radiography was conducted were not barricaded and posted at the 2 millirem per hour point.

Additionally, the responsible radiation protection representative did not walk down the area and ensure no unauthorized personnel were present prior to the conduct of radiography. Because this violation was of very low safety significance and has been entered into the licensees corrective action program as CR-GGN-2007-01582, it is being treated as a NCV, consistent with Section VI.A of the NRC Enforcement Policy:

NCV 05000416/2007002-03, Failure to Follow Procedural Guidance and Radiation Work Instructions While Supporting Radiography Operations.

.2 Failure to Follow Radiation Work Permit Instructions Prohibiting High Radiation Area

Entry

Introduction.

The inspector reviewed a self-revealing, NCV of TS 5.4.1 resulting from a failure to follow radiation work permit instructions prohibiting high radiation area entry.

The violation had very low safety significance.

Description.

On March 25, 2007, two outage workers entered into a high radiation area on the 139-foot elevation of the auxiliary steam tunnel. Dose rates in this area were as high as 125 millirems per hour at 30 centimenters from the source of radiation. The workers were using Radiation Work Permit 07-1518, Task 2, which controlled corrosion examinations and craft support and did not allow entry into high radiation areas. The

ENCLOSURE licensee was alerted to the entry into the high radiation area by the alarming dosimeter of one of the workers. Task 3 of the same radiation work permit allowed entry into high radiation areas and would have been the correct selection. However, Task 3 required workers to contact radiation protection personnel for high radiation area entry requirements. The workers acknowledged that they had not contacted radiation protection personnel for the required information. The licensee revoked the workers access to the radiological controlled area and the workers employer took disciplinary action because of the occurrence. The licensee documented the occurrence in their corrective action program.

Analysis.

The failure to follow radiation work permit instructions is a performance deficiency. 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 obtain information about potential radiological hazards could result in unplanned personnel dose. The occurrence involved the potential for unplanned, unintended dose resulting from actions contrary to licensee procedures and a radiation work permit which could have been significantly greater as a result of a single minor, reasonable alteration of the circumstances; therefore, the finding was evaluated using the Occupational Radiation Safety Significance Determination Process.

The finding was determined to be of 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 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> 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.

The finding was self-revealing because the licensee was alerted to the problem by an electronic dosimeter alarm and identification of the occurrence required no active and deliberate observation by the licensee.

Enforcement.

Technical Specification 5.4.1 requires procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Appendix A recommends, in Section 7, procedures for the control of radioactivity for limiting personnel exposure. EN-RP-100, Radworker Expectations, Revision 0, implements this requirement.

Procedure RP-100, Radiation Worker Expectations, Revision 0, Section 4, states that it is the responsibility of all individuals to know and follow radiation work permit instructions when performing radiological work. Radiation Work Permit 07-1518, Task 2, Worker Instruction 12, states, No high radiation or locked high radiation allow[ed] on this task.

Part 20.1003 of Title 10 of the Code of Federal Regulations, defines high radiation area as an area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 100 millirem (1 mSv) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 centimeters from the radiation source or 30 centimeters from any surface that the radiation penetrates. Radiation Work Permit 07-1518, Task 3, Worker Instruction 2, states, Contact RP for high radiation area entry requirements.

ENCLOSURE Radiation workers violated these requirements when two workers entered a high radiation area while working in accordance with Radiation Work Permit 07-1518, Task 2.

Further, the workers failed to contact radiation protection personnel to learn the entry requirements for the area, as required by Radiation Work Permit 07-1518, Task 3.

Because this violation was of very low safety significance and has been entered into the licensees corrective action program as CR-GGN-2007-01442, it is being treated as non-cited violation, consistent with Section VI.A of the NRC Enforcement Policy:

NCV 05000416/2007002-04, Failure to Follow Radiation Work Permit Instructions Prohibiting High Radiation Area Entry.

.3 Failure to Evaluate the Radiological Hazard Caused by Water Leaking in the Drywell

Introduction.

The inspector identified a Green NCV of 10 CFR 20.1501(a) because the licensee failed to evaluate the radiological hazard caused by water leaking from a valve in the drywell.

Description.

On March 19, 2007, during a tour of the drywell, an inspector observed two operators placing tags on accumulators on the 161-foot elevation. As a result of a leak in Valve B21F005, the operators had become wet with water from the reactor cooling system. The inspector promptly notified health physics personnel of the situation. The inspector also notified the radiation protection manager and outage control center personnel. The licensee had surveyed the area affected by the leaking water on the 161-foot elevation on March 18, 2007, and found contamination levels as high as 40,000 disintegrations per minute (beta/gamma). Despite the inspectors observation involving the operators, the next survey of the area was not conducted until March 21, 2007. This survey found contamination levels as high as 24 millirad per hour, a much higher contamination level. With radiological conditions changing this rapidly, the licensees survey frequency was not adequate to ensure current information was available to inform workers of potential radiological hazards.

The operators attempted to leave the radiological controlled area, but were unable to pass through the gamma-detecting personnel monitor without an alarm. The operators changed clothes and showered, but still were not able to pass through the monitors successfully. They were sent for whole body counts where small amounts of radioisotopes were detected. After comparing counts of the front and back sides of the individuals, radiation protection personnel concluded that the contamination was external and no committed effective dose equivalent was assigned.

Analysis.

The failure to perform a radiological survey of water leaking in the drywell is a performance deficiency. 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. Since this occurrence involves workers unplanned, unintended dose or potential of such a dose which could have been significantly greater as a result of a single minor, reasonable alteration of circumstances, this finding was evaluated with the Occupational Radiation Safety Significance Determination Process. The finding was determined to be of 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

ENCLOSURE 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 (Section 2OS1).

Enforcement.

Part 20.1501(a) of Title 10 of the Code of Federal Regulations, requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present.

Pursuant to 10 CFR 20.1003, a survey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation.

Part 20.1201(a) of Title 10 of the Code of Federal Regulations, states, in part, that the licensee shall control the occupational dose to individual adults to specified limits. The licensee violated 10 CFR 20.1501(a) when it failed to perform surveys frequently enough to ensure knowledge of changing radiological conditions and ensure personnel doses did not exceed regulatory limits. Because this failure to perform radiological surveys is of very low safety significance and has been entered into the licensees corrective action program as CR GGN-2007-1183 and CR-GGN-2007-1247, this violation is being treated as a NCV, consistent with Section VI.A of the NRC Enforcement Policy:

NCV 05000416/2007002-05, Failure to Evaluate the Radiological Hazard Caused by Water Leaking in the Drywell.

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual and collective radiation exposures ALARA. The inspector used the requirements in 10 CFR Part 20 and the licensees procedures required by TSs as criteria for determining compliance. The inspector interviewed licensee personnel and reviewed:

• Interfaces between operations, radiation protection, maintenance, maintenance planning, scheduling and engineering groups

• Integration of ALARA requirements into work procedure and radiation work permit documents

• Exposure tracking system

• Workers use of the low dose waiting areas

• First-line job supervisors contribution to ensuring work activities are conducted in a dose efficient manner

• Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas The inspector completed one of the required 15 samples and 5 of the optional samples.

ENCLOSURE

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

a. Inspection Scope

Initiating Events Cornerstone C

Unplanned Scrams Per 7,000 Critical Hours C

Unplanned Scrams With Loss Of Normal Heat Removal C

Unplanned Power Changes Per 7,000 Critical Hours Barrier Integrity Cornerstone C

Reactor Coolant System Leakage The inspectors sampled licensee submittals for the four performance indicators (PIs)listed above for the period from January through December 2005. The definitions and guidance of NEI 99-02, Regulatory Assessment Indicator Guideline, Revision 4, were used to verify the licensees basis for reporting each data element in order to verify the accuracy of PI data reported during the assessment period. The inspectors reviewed operator log entries, daily shift manager reports, plant computer data, condition reports, work orders, maintenance rule data, and PI data sheets to determine whether the licensee adequately reported the PIs listed above. Also, the inspectors interviewed the licensee personnel that were accountable for collecting and evaluating the PI data.

Occupational Radiation Safety Cornerstone Occupational Exposure Control Effectiveness The inspector reviewed licensee documents from June 1, 2006, through March 30, 2007.

The review included corrective action documentation that identified occurrences in locked high radiation areas (as defined in the licensees technical specifications), very high radiation areas (as defined in 10 CFR 20.1003), and unplanned personnel exposures (as defined in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Revision 4). Additional records reviewed included ALARA records and whole body counts of selected individual exposures. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the performance indicator data. In addition, the inspector toured plant areas to verify that high radiation, locked high radiation, and very high radiation areas were properly controlled. Performance indicator definitions and guidance contained in NEI 99-02, Revision 4, were used to verify the basis in reporting for each data element.

The inspector completed the required sample

(1) in this cornerstone.

ENCLOSURE Public Radiation Safety Cornerstone Radiological Effluent TS/Offsite Dose Calculation Manual Radiological Effluent Occurrences The inspector reviewed licensee documents from June 1, 2006, through March 30, 2007.

Licensee records reviewed included corrective action documentation that identified occurrences for liquid or gaseous effluent releases that exceeded performance indicator thresholds and those reported to the NRC. The inspector interviewed licensee personnel that were accountable for collecting and evaluating the performance indicator data.

Performance indicator definitions and guidance contained in NEI 99-02, Revision 4, were used to verify the basis in reporting for each data element.

The inspector completed the required sample

(1) in this cornerstone.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

The inspectors performed a daily screening of items entered into the licensee's CAP.

This assessment was accomplished by reviewing work orders and condition reports and attending corrective action review and work control meetings. The inspectors:

(1) verified that equipment, human performance, and program issues were being identified by the licensee at an appropriate threshold and that the issues were entered into the CAP;

(2) verified that corrective actions were commensurate with the significance of the issue; and

(3) identified conditions that might warrant additional follow-up through other baseline inspection procedures.

b. Findings

No findings of significance were identified.

.2 Selected Issue Follow-up Inspection

a. Inspection Scope

In addition to the routine review, the inspectors selected the three listed issues for more in-depth review. The inspectors considered the following during the review of the licensee's actions:

(1) complete and accurate identification of the problem in a timely manner;

(2) evaluation and disposition of operability/reportability issues;

(3) consideration of extent of condition, generic implications, common cause, and previous occurrences;

(4) classification and prioritization of the resolution of the problem;

(5) identification of root and contributing causes of the problem;

(6) identification of corrective actions; and

(7) completion of corrective actions in a timely manner.

ENCLOSURE C

CR-GGN-2006-4394, Inadvertent isolation of safeguards switchgear ventilation C

CR-GGN-2006-4458, Loose bolt in high pressure core spray pump breaker C

CR-GGN-2006-3501, Drywell purge compressor drain plug corrosion Documents reviewed by the inspectors are listed in the attachment.

b. Findings and Observations

Failure to Follow Procedure Resulting in Isolation of Switchgear Ventilation

Introduction.

A self-revealing Green NCV of TS 5.4.1(a) was identified for the failure to follow a surveillance procedure resulting in the isolation of ventilation to the Division 1 and Division 3 safety-related switchgear rooms.

Description.

The fire protection for the safeguard switchgear and battery rooms at Grand Gulf is provided by a carbon dioxide system. Upon actuation of the system, electro thermal links in the ventilation ducting melt, shutting the ventilation dampers to contain the carbon dioxide in the affected room.

On November 12, 2006, a nonlicensed operator and an electrician performed a surveillance test of the carbon dioxide fire suppression system involving a momentary actuation to verify carbon dioxide flow. To prevent a loss of room ventilation due to closure of the ventilation dampers, the procedure required the removal of the fuses for the electro thermal links. Although Step 5.4.1 of the procedure directed removal of a fuse from Panel TB1P64DXXX, the operator incorrectly directed the electrician to remove a fuse from Panel N1P64DXXX. The electrician did not notice that the panel was not the panel specified in the procedure. As a result, the electro thermal links melted and shut the ventilation dampers for each room tested. The operator tested the system in the Division 1 and Division 3 switchgear and battery rooms and the remote shutdown panel room before the electrician noticed the change in ventilation and stopped the surveillance.

In response to the loss of ventilation to the switchgear and battery rooms, operators declared the electrical systems operable based on engineering judgement due to the low outside air temperatures and took compensatory measures to open the doors to the affected rooms to provide additional cooling as necessary. Maintenance technicians replaced the affected electro thermal links and reopened the ventilation dampers. The ventilation system was restored to its original lineup approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> after the event began. The inspectors concluded no actual loss of operability occurred for the affected switchgear given the low outside air temperatures and the room temperatures during the duration of the event, as well as the available recovery methods, such as manually opening the shut dampers or the use of portable air blowers.

Analysis.

The failure to follow station procedures is a performance deficiency. The 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 and reliability 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.

ENCLOSURE 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.

Enforcement.

Technical Specification 5.4.1(a) requires written procedures to be implemented as recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Appendix A recommends procedures governing maintenance practices.

I of Procedure 06-OP-SP64-R-0002, 10 Ton CO2 Systems Puff Test, Revision 105 directs operators to remove the fuses for the associated carbon dioxide system prior to conducting the system test. Contrary to this requirement, plant operators failed to remove the correct fuses prior to performing the surveillance test on November 12, 2006. Because this violation was of very low safety significance and was entered in the corrective action program as CR-GGN-2006-4394, this violation is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy:

NCV 05000416/2007002-06, Failure to Follow Procedure Resulting in Isolation of Switchgear Room Ventilation.

Failure to Identify and Correct Standby Service Water System Leakage

Introduction.

The inspectors identified a Green NCV 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.

Description.

The drywell purge compressors at Grand Gulf pump air from the containment through the drywell post-accident to provide drywell vacuum relief, reduce post accident radiation levels, and to prevent the concentration of hydrogen inside the drywell. The purge compressors are cooled by standby service water via the compressor oil coolers.

During a containment walkdown, the inspectors discovered significant corrosion around a drain plug on the service water side of the Train A purge compressor oil cooler. The corrosion had progressed to the point that a portion of the drain plug had rusted off below the cooler surface and was resting in the drain pan below the cooler. A failure analysis conducted later by licensee engineers showed that the service water leakage from a complete failure of the drain plug would have rendered the Division I standby service water system inoperable.

The licensee determined the corrosion was caused by leakage coming from a leakoff line installed inside the drain plug. The drain plugs were made of zinc and were designed by the vendor as sacrificial anodes to provide corrosion protection for the oil cooler.

Leakage through the leakoff lines was an indication that the sacrificial plug required replacement. The licensee did not have a preventive maintenance program or mechanism to monitor the condition of the drain plugs. In fact, licensee engineers performing a nondestructive examination of the oil cooler noted that the corroded area had been painted over at some point in the past.

When replacing a cooler end cover on the Train B purge compressor in September 2006, the licensee initiated Condition Report CR-GGN-2006-3501 documenting corrosion on the zinc drain plugs. As corrective action, the licensee performed an engineering modification to replace the drain plugs with plugs made from carbon steel. The licensee

ENCLOSURE did not evaluate the condition of the drain plugs on the Train A purge compressor oil cooler or replace the two other drain plugs on the Train B oil cooler. The inspectors concluded the licensees inadequate extent of condition evaluation led to the failure to identify a condition adverse to quality.

As part of the corrective actions in Condition Report CR-GGN-2006-4762 written in response to this issue, the licensee replaced the zinc drain plugs with carbon steel plugs to provide an adequate standby service water pressure boundary.

Analysis.

The failure to correct a condition adverse to quality is a performance deficiency. This finding is more than minor because if left uncorrected, the zinc drain plugs could have deteriorated to a point at which standby 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 to prevent undesirable consequences. 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.

Enforcement.

Criterion XVI, Corrective Action, of Appendix B to 10 CFR Part 50 requires, in part, that conditions adverse to quality are promptly identified and corrected.

Contrary to this requirement, the licensee failed to identify and correct leakage from the Train A drywell purge compressor oil cooler drain plug. Because this violation was of very low safety significance and was entered in the corrective action program as CR-GGN-2006-4762, this violation is being treated as a NCV consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000416/2007002-07, Failure to Identify and Correct Standby Service Water System Leakage.

4OA3 Event Follow-up

.1 Personnel Performance During Nonroutine Evolutions, Events, and Transients

a. Inspection Scope

The inspectors:

(1) reviewed operator logs, plant computer data, and/or strip charts for the below listed evolutions to evaluate operator performance in coping with nonroutine events and transients;

(2) verified that operator actions were in accordance with the response required by plant procedures and training; and

(3) verified that the licensee has identified and implemented appropriate corrective actions associated with personnel performance problems that occurred during the events sampled.

• On March 16, 2007, the inspectors observed operator performance in the control room during a planned, nonroutine plant power reduction to sixty-five percent rated thermal power to perform control rod stroke time testing in advance of refueling outage RF15. The inspectors observed operators controlling this

ENCLOSURE evolution in accordance with Integrated Operating Instruction 03-1-01-2, Power Operations, Revision 132, and Equipment Performance Instruction 04-S-03-C11-5, Control Rod Stroke Time Testing, Revision 109.

The inspectors completed one sample.

b. Findings

No findings of significance were identified.

.2 (Closed) LER 05000416/2006-001-00: Division 1 Diesel Generator Exhaust Valve Failure

On May 11, 2006, the Division I emergency diesel generator tripped as a result of the failure of an exhaust valve in Cylinder 8L during a postmaintenance test. This event and its regulatory aspects are the subject of NRC Special Inspection Report 05000416/2006-010. The LER was reviewed by the inspectors and no new findings of significance were identified. The licensee documented this issue in CR-GGN-2006-1955. This LER is closed.

4OA6 Meetings, Including Exit

On January 24, 2007, the operations inspector discussed operator requalification inspection results with Mr. R. Collins, Operations Manager, and other members of the licensees staff at the conclusion of the inspection. The licensee acknowledged the findings presented in the exit meeting. The inspector verified that no proprietary information was reviewed during the inspection.

On January 25, 2007, the reactor inspector presented the maintenance effectiveness inspection results to Mr. M. Krupa, Acting General Manager, and other members of licensee management at the conclusion of the onsite inspection. The inspector verified that no proprietary information was reviewed during the inspection.

On March 30, 2007, the health physics inspector presented the occupational radiation safety inspection results to Mr. D. Wiles, Director, Engineering, and other members of the licensees staff who acknowledged the findings. The inspector confirmed that proprietary information was not provided or examined during the inspection.

On April 3, 2007, the resident inspectors presented the inspection results to Mr. R. Brian and others who acknowledged the findings. Proprietary information was reviewed by the inspectors and was returned to the licensee at the end of the inspection.

4OA7 Licensee-Identified Violations

The following violations of very low significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a NCV.

• Technical Specification 5.4.1(a) requires written procedures to be implemented as recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Appendix A recommends procedures governing surveillance testing.

Section 6.8.8 of Administrative Procedure 01-S-06-12, GGNS Surveillance

ENCLOSURE Program, Revision 109, requires supervisors to review surveillance test results for technical accuracy and adequacy. During an operating experience review conducted on January 15, 2007, the licensee determined that a station battery intercell resistance measurement surveillance conducted per TS Surveillance Requirement 3.8.4.5 on August 9, 2005, had been performed incorrectly resulting in inaccurate test data. The data collected, though within the limits of the TS, were four to five times higher than values recorded in previous tests, and if correct would have indicated substantial internal degradation of the Division 2 vital battery. The inaccurate data was not discovered during supervisory review at the time of the surveillance. Subsequent performance of the surveillance test on January 15, 2007 confirmed that the previous results were inaccurate. This event was documented in the corrective action program as Condition Report CR-GGN-2007-0174. This finding is of very low safety significance since there was no actual loss of operability.

• Technical Specification 5.4.1(a) requires written procedures to be implemented as recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Appendix A recommends procedures for performing maintenance. Section 7.1.9 of Procedure 07-S-12-61, Inspection of GE Magne Blast Circuit Breakers, Revision 3, directs maintenance technicians to ensure all fasteners are tight to ensure there are no parts that could become dislodged within the breaker. During a breaker inspection conducted as part of a planned system outage on November 16, 2006, GGNS maintenance technicians discovered that a bolt had fallen out of the control block of the high pressure core spray pump breaker and was resting unrestrained inside the breaker cubicle. This event was documented in the corrective action program as Condition Report CR-GGN-2006-4458. This finding is of very low safety significance since there was no actual loss of operability.

• Technical Specification 5.4.1(a) requires written procedures to be implemented as recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Appendix A recommends procedures governing surveillance testing.

Section 5.1.4b of Surveillance Procedure 06-RE-1B33-D-0001, Jetpump Functional Test, Revision 108, requires operators to contact reactor engineering for evaluation of jetpump performance in the event the acceptance criteria of the procedure are not met. On March 15, 2007, an operator implementing the daily surveillance procedure noted that the acceptance criteria for one jetpump had not been met for the previous four performances of the surveillance procedure, but reactor engineering had not been notified to evaluate the condition of the jetpump as required. Subsequent evaluation by reactor engineering determined that one of the five nozzles on the jetpump was partially plugged. This event was documented in the corrective action program as Condition Reports CR-GGN-2007-1061 and CR-GGN-2007-1071. This finding is of very low safety significance since there was no actual loss of operability.

• Criterion V, Instructions, Procedures and Drawings, of Appendix B to 10 CFR Part 50 states, in part, that activities affecting quality shall be prescribed by documented instructions and shall be accomplished in accordance with those instructions. Attachment 9.1 of Procedure ENS-DC-118, ER Response Closure, Revision 4, requires drawings to be updated following the installation of plant modifications. On March 27, 2007, licensee engineers developed modification

ENCLOSURE ER-2007-0039 to install additional support plates for the moisture separator in the upper containment pool using a drawing that had not been updated to reflect modification ER-2001-0250 installed in July 2002, resulting in support plates that were too small for the intended purpose. The licensee discovered the support plates were improperly sized after implementing the modification while the separator was suspended above the plates. This event was documented in the corrective action program as Condition Report CR-GGN-2007-1295. This finding is of very low safety significance since it did not result in an actual loss of safety function.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

C. Abbott, Acting Manager, Quality Assurance

W. Abraham, Sr. Engineering Associate

C. Bottemiller, Manager, Plant Licensing

R. Brian, Vice President, Operations

M. Causey, Senior Lead Technical Specialist

K. Christian, Engineering Code Programs Supervisor

R. Collins, Manager, Operations

D. Coulter, Licensing Specialist, Plant Licensing

T. Curtis, Supervisor, Radiation Protection

M. Cross, ISI Coordinator, Grand Gulf

L. Eaton, Senior Lead Engineer

N. Edney II, Supervisor, Radiation Protection

C. Ellsaesser, Manager, Planning and Scheduling

M. Guynn, Manager, Emergency Preparedness

E. Harris, Manager, Corrective Action and Audits

M. Krupa, Director, Nuclear Safety Assurance

M. Larson, Senior Licensing Engineer

D. Mcdirmid, Technical Specialist IV

C. Mason, Quality Assurance Auditor

T. Tankersley, Manager, Training

T. Thornton, Manager Design Engineering

J. Reed, General Manager, Plant Operations

J. Robertson, Manager, Refueling Services

M. Rohrer, Manager, System Engineering

S. Scott, Central Engineering

R. Sumrall, Emergency Planner

T. Tankersley, Manager, Training

W. Trichell, Supervisor, Radiation Protection

K. Walker, Manager, Reactor Engineering

D. Wiles, Director, Engineering

D. Wilson, Supervisor, Design Engineering

R. Wilson, Superintendent, Radiation Protection

P. Worthington, Supervisor, Engineering

NRC personnel

W. Walker, Senior Project Engineer, Reactor Project Branch C

R. Bywater, Senior Reactor Analyst, Region IV

ATTACHMENT

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Opened and Closed

05000416/2007002-01 NCV Failure to Follow Procedures Resulting in an Inadequate Operability Evaluation (Section 1R15)

05000416/2007002-02 NCV Ineffective Command and Control Results in Inappropriate Valve Manipulations (Section 1R22)

05000416/2007002-03 NCV Failure to Follow Procedural Guidance and Radiation Work Instructions While Supporting Radiography Operations (Section 2OS1)

05000416-2007002-04 NCV Failure to Follow Radiation Work Permit Instructions Prohibiting High Radiation Area Entry (Section 2OS1)

05000416/2007002-05 NCV Failure to Evaluate the Radiological Hazard Caused by Water Leaking in the Drywell (Section 2OS1)

05000416/2007002-06 NCV Failure to Follow Procedure Resulting in Isolation of Switchgear Room Ventilation (Section 4OA2)

05000416/2007002-07 NCV Failure to Identify and Correct Standby Service Water System Leakage (Section 4OA2)

Closed

05000416/2006-001-00 LER Division 1 Diesel Generator Exhaust Valve Failure

Discussed

None