Text

NRC Integrated Inspection Report Number 05000416-10-005 and Notice of Violation
	ML110420273
Person / Time
Site:	Grand Gulf
Issue date:	02/10/2011
From:	Vincent Gaddy; NRC/RGN-IV/DRP/RPB-C
To:	Douet J; Entergy Operations
References
	EA-10-256
	Download: ML110420273 (76)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION REGI ON I V 612 EAST LAMAR BLVD, SUITE 400 ARLINGTON, TEXAS 76011-4125 February 10, 2011 EA-10-256 James Vice President Operations Entergy Operations, Inc.

Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150 Subject: GRAND GULF - NRC INTEGRATED INSPECTION REPORT NUMBER 05000416/2010005 AND NOTICE OF VIOLATION

Dear Mr. Douet:

On December 31, 2010, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Grand Gulf Nuclear Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on January 11, 2011, with Jeremy Browning, General Plant Manager, and other members of your staff.

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

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

Based on the results of this inspection, the NRC has identified an issue that was evaluated under the risk significance determination process as having very low safety significance (Green). The NRC has also determined that a violation is associated with this issue. The violation was evaluated in accordance with the NRC Enforcement Policy. The current Enforcement Policy is included on the NRC's website at (http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html).

The violation is cited in the enclosed Notice of Violation and the circumstances surrounding it is described in detail in the subject inspection report. The violation is being cited in the Notice because the licensee failed to restore compliance with NRC requirements within a reasonable time after July 30, 2009. This is consistent with the NRC Enforcement Policy; Section 2.3.2, which states, in part, that a cited violation will be considered if the licensee fails to restore compliance within a reasonable time after a violation is identified.

EA-10-256 Entergy Operations, Inc. -2-The NRC has concluded that information regarding the reason for the violation, the corrective actions taken to prevent recurrence and the date when full compliance was achieved is already adequately addressed in this inspection report. Therefore, you are not required to respond to this letter unless the description herein does not accurately reflect your corrective actions or your position. In that case, or if you choose to provide additional information, you should follow the instructions specified in the enclosed Notice.

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

Additionally, three licensee-identified violations, which were 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 a noncited violations, consistent with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the violation or the significance of the noncited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C.

20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Grand Gulf Nuclear Station. In addition, if you disagree with the crosscutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector at Grand Gulf Nuclear Station.

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

Sincerely,

/RA/

Vincent Gaddy, Chief Project Branch C Division of Reactor Projects Docket: 50-416 License: NPF-29

EA-10-256 Entergy Operations, Inc. -3-Enclosure 1: Notice of Violation Enclosure 2: NRC Inspection Report 05000416/2010005

Attachment:

Supplemental Information

REGION IV==

Docket: 05000416 License: NPF-29 Report: 05000416/2010005 Licensee: Entergy Operations, Inc.

Facility: Grand Gulf Nuclear Station Location: 7003 Baldhill Road Port Gibson, MS 39150 Dates: September 28, 2010 through December 31, 2010 Inspectors: R. Smith, Senior Resident Inspector A. Barrett, Acting Senior Resident Inspector T. Burns, Senior Reactor Inspector, Region I J. Draper, Acting Resident Inspector P. Elkman, Acting Resident Inspector G Guerra, CHP, Emergency Preparedness Inspector B. Hagar, Senior Project Engineer S. Hedger, Operations Engineer D. Jones, Senior Reactor Inspector, Region III R. Latta, Senior Reactor Inspector, Region IV (Team Leader)

R. Kumana, Project Engineer D. Norwood, Acting Resident Inspector D. Overland, Resident Inspector (Waterford)

P. Prescott, Senior Quality and Vendor Program Engineer, NRR Approved By: Vincent Gaddy, Chief, Project Branch C Division of Reactor Projects-1- Enclosure 2

SUMMARY OF FINDINGS

IR 05000416/2010005; 01/01/2010 - 12/31/2010; Grand Gulf Nuclear Station, Integrated

Resident and Regional Report; Operability Evaluations, Surveillance Testing, Correction of Emergency Preparedness Weaknesses and Deficiencies and Identification and Resolution of Problems.

The report covered a 3-month period of inspection by resident inspectors and announced baseline inspections by region-based inspectors. Additionally, the report covered a 12-month period of inspection by a Region IV inspector, plus a one-week inspection by inspectors from NRR and Regions I, III, and IV. Five Green noncited violations of significance were identified and one Green cited violation of significance was identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609,

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

Revision 4, dated December 2006.

NRC-Identified Findings and Self-Revealing Findings

Cornerstone: Initiating Events

Green.

The inspectors identified a noncited violation of 10 CFR 50 Appendix B, Criterion V, for an inadequate shutdown procedure resulting in power and level oscillations in the reactor. The revised procedure failed to require the plant to be placed in startup feedwater level control during low power operations. The performance deficiency was self-revealing; however the inspectors added significant value by identifying inadequate condition report classification, causal evaluation, and corrective actions. As corrective action, the licensee planned to revise the procedure to ensure the plant is placed in startup feedwater control during low power operations. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-05140.

The finding is more than minor because it was associated with the initiating events cornerstone attribute of procedure quality and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Phase 1 Screening Worksheet in attachment 4 of Inspection Manual 0609, Significance Determination Process, the inspectors determined that the finding had very-low safety significance (Green) because in Table 4a of the Phase 1 Screening Worksheet, the finding was a transient initiator that did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. This finding has a crosscutting aspect in the area of human performance associated with the decision-making component, because station management failed to use conservative assumptions to demonstrate that the change to the shutdown operating procedure was safe prior to proceeding H.1(b) (Section 4OA2).

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of 10 CFR 50 Appendix B,

Criterion V, involving a failure to follow procedures, which resulted in an inadequate operability evaluation. On December 5, 2010, a spurious actuation of the standby service water pump house ventilation system occurred, resulting in the pump house temperatures dropping below the design limit. In their operability evaluation, the licensee failed to consider the impact of the actual freezing conditions occurring at the site at that time, and operations did not secure the fan after the spurious actuation until questioned by the inspectors.

The licensee subsequently revised the operability evaluation to properly account for environmental conditions in the pump house. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2011-00151.

This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance as it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Phase 1 Worksheet in attachment 4 of Manual Chapter 0609, Significance Determination Process, the inspectors determined that this finding was of very-low safety significance (Green) because in Table 4a of the Phase 1 Screening Worksheet, all of the questions for the Mitigating Cornerstone were answered in the negative. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective actions component because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of condition

P.1(c) (Section 1R15).

Green.

The inspectors reviewed a self-revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, associated with the licensees failure to take timely corrective actions to correct a condition adverse to quality associated with degradation of the Reactor Core Isolation Cooling (RCIC) flow control system, which ultimately resulted in the RCIC turbine governor failing its surveillance test. On September 23, 2010, the licensee replaced the electric governor-magnetic pickup, correcting the condition adverse to quality. The licensee entered this issue into the corrective action program as Condition Report CR-GGN-2010-06850.

This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance as it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding affected the mitigating systems cornerstone because the deficiency degraded an operating margin associated with the short term heat removal capability of the RCIC system. Using the Phase 1 Screening Worksheet in attachment 4 of Inspection Manual 0609, Significance Determination Process, the inspectors determined that the finding had very-low safety significance (Green) because in Table 4a of the Phase 1 Screening Worksheet, all of the questions for the Mitigating Cornerstone were answered in the negative. This finding has a cross-cutting aspect in the resources component of the Human Performance area because the licensee did not maintain long-term plant safety by maintaining design margins

H.2(a) (Section 1R22).

Green.

Inspectors identified a noncited violation of 10 CFR 50, Appendix B,

Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. The licensee performed extensive reviews, and inspectors performed independent reviews of the licensees conclusions as well as independent sampling, to confirm that improper or missed inspections did not actually affect the operability of plant equipment. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective action program under Condition Reports CR-HQN 2009-01184 and CR-HQN-2010-0013.

The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the Design Control attribute of the Mitigating Systems cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to have very low safety significance in Phase 1 of Manual Chapter 0609, Significance Determination Process, since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this performance deficiency involved a cross-cutting aspect related to the human performance in decision-making (H.1a), because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate H.1(a) (Section 4OA2).

Green.

Inspectors identified a noncited violation of 10 CFR 50, Appendix B,

Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of

Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program. Specifically, the individual assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as Condition Report CR-HQN-2010-00386. Failure to ensure that an individual assigned to the position Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing Significance Determination Process guidance, so it was determined to be of very low safety significance using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no cross-cutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago (Section 4OA2).

Cornerstone: Emergency Preparedness

Green.

A cited violation of 10 CFR 50.47(b)(10) was identified for failure to develop and have in place guidelines for the choice of protective actions during an emergency that were consistent with federal guidance. Federal guidance for the choice of protective actions during an emergency is described in EPA-400-R-92-001 and states, in part, that evacuation is seldom justified when doses are less than protective action guides. The licensees automatic process that extended existing protective action recommendations with changes in wind direction without considering radiation dose was identified as a performance deficiency.

This finding is more than minor because it affects the Emergency Preparedness Cornerstone objective of implementing adequate measures to protect the health and safety of the public during a radiological emergency, and is associated with the cornerstone attributes of emergency response organization performance and procedure quality. This finding was determined to be of very low safety significance because it was a failure to comply with NRC requirements, was associated with risk significant planning standard 10 CFR 50.47(b)(10), and was not a risk significant planning standard functional failure or a planning standard degraded function. The finding was not a functional failure or degraded planning standard function because appropriate protective action recommendations for the public would have been made for all areas where protective action guides were exceeded. This finding is a cited violation of 10 CFR 50.47(b)(10) because the licensee failed to restore compliance with NRC requirements in a timely manner.

The finding is related to the corrective action element of the problem identification and resolution crosscutting aspect because the licensee failed to take corrective actions to address the safety issue in a timely manner P.1(d) (Section 1EP5).

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 licensees corrective action program. These violations and corrective action tracking numbers (condition report numbers) are listed in Section 4OA7.

REPORT DETAILS

Summary of Plant Status

Grand Gulf Nuclear Station began the inspection period at full rated thermal power. On October 14, 2010, operators reduced power to 90 percent for monthly control rod testing. The plant was returned to 100 percent power on October 15, 2010. On November 4, 2010, operators reduced reactor power to 67 percent for a planned control rod sequence exchange and scram time testing. The plant returned to 100 percent power on November 5, 2010. On December 10, 2010, operators reduced power to 90 percent for monthly control rod testing. The plant was returned to 100 percent power on the same day. On December 14, 2010, power was reduced to 80 percent to evaluate steam leaks in the fifth stage B feedwater heater room. The licensee back seated a steam supply valve to stop one leak, and they also determined the location of another leak to be on the welded pipe fitting connection on the 24-inch extraction steam piping at the high point vent a 3/4-inch line. The plant was returned to 100 percent power on December 15, 2010. On December 19, 2010, power was reduced to 80 percent to perform a repair of the steam leak on the welded pipe fitting connection on the fifth stage B feedwater heater. The licensee had to back out of the repair effort due to high room temperatures, and the plant was returned to 100 percent power the same day. On December 24, 2010, power was reduced to 80 percent to repair the steam leak on the welded pipe fitting connection on fifth stage B feedwater heater. After the leak was repaired, the plant was returned to 100 percent power on the same day. The plant remained at full rated thermal power for the remainder of the inspection period.

REACTOR SAFETY

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

1R04 Equipment Alignments

Main article: IP 71111.04

.1 Partial Walkdown

a. Inspection Scope

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

Reactor core isolation cooling following system maintenance (note smart sample 2009-2 completed)

Division 2 drywell and containment hydrogen analyzers while division 1 was out of service for unplanned maintenance

Noble metal chemistry mitigation monitoring system following system maintenance

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

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

b. Findings

No findings were identified.

.2 Complete Walkdown

a. Inspection Scope

On October 25, 2010, the inspectors performed a complete system alignment inspection of the low pressure core spray system to verify the functional capability of the system.

The inspectors selected this system because it was considered both safety significant and risk significant in the licensees probabilistic risk assessment. The inspectors inspected the system to review mechanical and electrical equipment line ups, electrical power availability, system pressure and temperature indications, as appropriate, component labeling, component lubrication, component and equipment cooling, hangers and supports, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. The inspectors reviewed a sample of past and outstanding work orders to determine whether any deficiencies significantly affected the system function. In addition, the inspectors reviewed the corrective action program database to ensure that system equipment-alignment problems were being identified and appropriately resolved. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one complete system walkdown sample as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings were identified.

1R05 Fire Protection

Main article: IP 71111.05

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

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

Lower cable spreading room (OC402)

Upper cable spreading room (OC 712)

Motor control center (1A410)

Snubber test facility control room and test room (1A519A & 1A519B)

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

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

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

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

b. Findings

No findings were identified.

1R06 Flood Protection Measures

Main article: IP 71111.06

a. Inspection Scope

The inspectors reviewed the UFSAR, the flooding analysis, and plant procedures to assess susceptibilities involving internal flooding; reviewed the corrective action program to determine if licensee personnel identified and corrected flooding problems; inspected underground bunkers/manholes to verify the adequacy of sump pumps, level alarm circuits, cable splices subject to submergence, and drainage for bunkers/manholes; and verified that operator actions for coping with flooding can reasonably achieve the desired outcomes. The inspectors also inspected the areas listed below to verify the adequacy of equipment seals located below the flood line, floor and wall penetration seals, watertight door seals, common drain lines and sumps, sump pumps, level alarms, and control circuits, and temporary or removable flood barriers. Specific documents reviewed during this inspection are listed in the attachment.

November 18, 2010, control building These activities constitute completion of one flood protection measures inspection sample as defined in Inspection Procedure 71111.06-05.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program

Main article: IP 71111.11

.1 Quarterly Inspection

a. Inspection Scope

On September 30, 2010, the inspectors observed a crew of licensed operators in the plants simulator to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas:

Licensed operator performance

Crews clarity and formality of communications

Crews ability to take timely actions in the conservative direction

Crews prioritization, interpretation, and verification of annunciator alarms

Crews correct use and implementation of abnormal and emergency procedures

Control board manipulations

Oversight and direction from supervisors

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

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

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

b. Findings

No findings were identified.

.2 Annual Inspection

a. Inspection Scope

The inspectors reviewed the annual operating examination test results for 2010. Since this was the first half of the biennial requalification cycle, the licensee was not required to administer a written examination. These results were assessed to determine if they were consistent with NUREG 1021, "Operator Licensing Examination Standards for Power Reactors," guidance and Inspection Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance Significance Determination Process," requirements.

This review included the test results for a total of eight crews composed of 50 licensed operators, which included: 15 shift-standing senior operators, 15 staff senior operators, 15 shift-standing reactor operators, and five staff reactor operators. There was one crew failure and no individual failures on the simulator. The failures were remediated following the exam, and the crew passed a simulator re-examination prior to resuming watch standing. Corrective actions being evaluated as a result of the crew failure are documented in Condition Report CR-GGN-2010-07326.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12

a. Inspection Scope

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

Reactor recirculation system (B33)

High pressure core spray diesel generator (P81)

Feedwater and plant service water systems

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

Implementing appropriate work practices

Identifying and addressing common cause failures

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

Characterizing system reliability issues for performance

Charging unavailability for performance

Trending key parameters for condition monitoring

Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or -(a)(2)

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

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

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

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

Main article: IP 71111.13

a. Inspection Scope

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

Radial well ventilation system

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

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

b. Findings

No findings were identified.

1R15 Operability Evaluations

Main article: IP 71111.15

a. Inspection Scope

The inspectors reviewed the following issues:

Spent fuel pool boron density calculations non-conservative

Freon leaks discovered in the division 2 control room air-conditioning system

Combustion air inlet damper design not consistent with vendor manual drawings

Reactor recirculation flow control valve drift

Standby service water pump house ventilation fan auto-start The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that technical specification operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the technical specifications and UFSAR to the licensee personnels evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the

inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Specific documents reviewed during this inspection are listed in the attachment.

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

b. Findings

Introduction.

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

Description.

The standby service water pumps and supporting systems are installed in pump houses to protect the systems from adverse external conditions and events. On December 5, 2010, a spurious actuation of the standby service water pump house ventilation system caused a fan to start. In response, the licensee initiated a condition report and performed an operability evaluation which stated, in part, that the actuation of the fan would not impact the function of the standby service water system. The next morning, the inspectors reviewed the subject operability evaluation and found that the licensee had failed to consider the impact of the actual freezing environmental conditions that had been occurring at the site at that time, and that the operations staff had not secured the fan after the spurious actuation, which had resulted in the pump-house temperatures dropping below the design limit of 50 degrees Fahrenheit. After the inspectors questioned the control room staff about continued operation of the pump house ventilation fan, operators secured the fan. Further review by the licensee showed that temperature detectors in the pump house had reached a low of 48.7 degrees Fahrenheit while the fans were operating. Plant personnel provided a calculation showing that the lowest temperature that could impact equipment function would have been 33.45 degrees Fahrenheit. The licensee subsequently revised the operability evaluation to properly account for environmental conditions in the pump house. The licensee entered the inspectors findings into their corrective action program as Condition Report CR-GGN-2010-08515.

Analysis.

The licensees failure to perform an adequate operability evaluation in accordance with station procedures is a performance deficiency. This performance deficiency is more than minor and is therefore a finding because it is associated with the mitigating systems cornerstone attribute of equipment performance and it adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, this finding resulted in significantly degrading the margin between environmental temperature and the temperature at which component functions would have been affected. Using the Phase 1 Worksheet in attachment 4 of Manual Chapter 0609, Significance Determination Process, the inspectors determined that this finding was of very-low safety significance (Green) because in Table 4a of the Phase 1 Screening Worksheet, all of the questions for the Mitigating Cornerstone were answered

in the negative. The inspectors determined that the apparent cause of this finding was that licensee personnel had failed to thoroughly evaluate the impact of the spurious actuation of the standby service water pump house ventilation fan, and that this cause reflects current licensee performance. Therefore, this finding has a crosscutting aspect in the area of problem identification and resolution associated with corrective actions because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions P.1(c).

Enforcement.

10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions. Section 5.4[4] of EN-OP-104, Operability Determinations, Revision 4, is a documented instruction that required licensee personnel who prepare operability evaluations to evaluate the applicable system design specifications and requirements for environmental qualification of components. Contrary to the above, on December 5, 2010, when licensee personnel completed an operability evaluation for the spurious actuation of the standby service water pump house ventilation fan, they failed to evaluate the applicable system design specifications and requirements for environmental qualification of components, in that they failed to evaluate the capability of components in the standby service water pump house to withstand the environmental conditions that were occurring at the site at that time.

Because this violation was of very-low safety significance and was entered in the corrective action program as Condition Report CR-GGN-2011-00151, this violation is being treated as a noncited violation consistent with Section 2.3 of the NRC Enforcement Policy, and has been designated NCV 05000416/2010005-01, Inadequate Operability Evaluation Following a Spurious Actuation of the Standby Service Water Pump House Ventilation Fan.

1R19 Postmaintenance Testing

Main article: IP 71111.19

a. Inspection Scope

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

Power factor testing of station transformer 11 following transformer outage The inspectors selected these activities based upon the structure, system, or component's ability to affect risk. The inspectors evaluated these activities for the following:

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

Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate

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

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

b. Findings

No findings were identified.

1R22 Surveillance Testing

Main article: IP 71111.22

a. Inspection Scope

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

Preconditioning

Evaluation of testing impact on the plant

Acceptance criteria

Test equipment

Procedures

Jumper/lifted lead controls

Test data

Testing frequency and method demonstrated technical specification operability

Test equipment removal

Restoration of plant systems

Fulfillment of ASME Code requirements

Updating of performance indicator data

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

Reference setting data

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

October 22, 2010, reactor core isolation cooling surveillance test

December 15, 2010, low pressure coolant injection C quarterly pump inservice test Specific documents reviewed during this inspection are listed in the attachment.

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

b. Findings

Introduction.

The inspectors reviewed a self-revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, for failure to take timely corrective actions to correct a condition adverse to quality associated with degradation of the reactor core isolation cooling (RCIC) flow control system.

Description.

On October 26, 2009, during a monthly test of the RCIC system, the system met the acceptance criteria for RCIC turbine inlet pressure and pump speed, flow rate, and discharge pressure. However, during these tests, control room operators noticed unexpected performance of the RCIC flow controller. The controller which previously had a demand signal of 70-80 percent open to achieve the required discharge pressure during the test had a demand signal of greater than 100 percent open. In response, the licensee initiated Condition Report CR-GGN-2009-05640 and Work Order (WO) 215669 to troubleshoot the controller. In WO 215669, the work instruction text indicated that the work should be performed before the next quarterly test of the RCIC system, which was scheduled for November 11, 2009. WO 215669 was assigned a priority of 4G, which directed the work planner to schedule the work as resources allow within the normal process. The condition report documented a partial causal evaluation that noted that the RCIC pump was no longer achieving the expected 80 psig differential pressure to enable injection into the reactor vessel. The evaluation noted that this margin had been reduced to 5 psig. The final causal evaluation was not

completed, with a justification that data obtained from WO 215669 was needed to complete the analysis.

Because WO 215669 had a priority of 4G, emergent issues and manpower limitations prompted the licensee to delay its completion first to February 8, 2010; then to March 10, 2010; June 20, 2010; and finally to the next scheduled quarterly RCIC test scheduled for September 23, 2010. Because the ACE was on hold pending completion of the WO, the ACE was also similarly rescheduled.

On September 21, 2010, a few days before the scheduled quarterly RCIC test, the licensee used WO 215669 to adjust the null voltage of the RCIC flow controller, in an effort to correct the degraded condition. However, following this adjustment, the RCIC failed the test because it was not able to achieve the required flow rate and discharge pressure. In response, the licensee initiated ACE CR-GGN-2010-06850 to evaluate the test failure. In this evaluation, the licensee identified a deficiency in the pumps electric governor-magnetic pickup (EGM). After the licensee replaced the EGM and the flow controller, a post-maintenance test on September 23, 2010, revealed that the RCIC was able to produce the required flow rate and discharge pressure, and that the discharge pressure had increased such that the difference between the discharge pressure of RCIC and reactor pressure had been restored to the normal value of approximately 80 psig. Therefore, the licensee resolved on September 23, 2010, a condition adverse to quality that had been identified on October 26, 2009.

Analysis.

The failure to take timely corrective action to address an adverse condition in the RCIC flow control system is a performance deficiency. This performance deficiency is more than minor because it is associated with the mitigating systems cornerstone attribute of equipment performance and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Phase 1 Screening Worksheet in attachment 4 of Inspection Manual 0609, Significance Determination Process, the inspectors determined that the finding had very-low safety significance (Green) because in Table 4a of the Phase 1 Screening Worksheet, all of the questions for the Mitigating Cornerstone were answered in the negative. The inspectors determined that the apparent cause of this finding was the licensees failure to properly prioritize WO 215669 (the work order associated with correcting the degraded condition with the RCIC flow control system), and that this apparent cause is reflective of current licensee performance. This finding therefore has a cross-cutting aspect in the resources component of the Human Performance area because the licensee did not maintain long-term plant safety by maintaining design margins H.2(a).

Enforcement.

10 CFR Part 50 Appendix B, Criterion XVI requires, in part, that conditions adverse to quality are promptly identified and corrected. Contrary to the above, the licensee did not promptly correct a condition adverse to quality, in that the licensee identified degraded RCIC pump discharge pressure as a condition adverse to quality on October 26, 2009, but did not correct that condition until September 23, 2010.

Because this violation was of very-low safety significance and the licensee entered it into their corrective action program as condition report CR-GGN-2010-06850, this violation is

being treated as a noncited violation, consistent with Section 2.3 of the NRC Enforcement Policy, and has been designated NCV 05000416/2010005-02, Untimely Corrective Actions in Response to Deficiencies in the RCIC Flow Control System.

Cornerstone: Emergency Preparedness

1EP2 Alert Notification System Testing

Main article: IP 71114.02

a. Inspection Scope

The inspector discussed with licensee staff the operability of offsite siren emergency warning systems, and backup alerting methods, to determine the adequacy of licensee methods for testing the alert and notification system in accordance with 10 CFR Part 50, Appendix E. The licensees alert and notification system testing program was compared with criteria in NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1; FEMA Report REP-10, Guide for the Evaluation of Alert and Notification Systems for Nuclear Power Plants; and the licensees current FEMA-approved alert and notification system design report. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one sample as defined in Inspection Procedure 71114.02-05.

b. Findings

No findings were identified.

1EP3 Emergency Response Organization Augmentation Testing

Main article: IP 71114.03

a. Inspection Scope

The inspector discussed with licensee staff the operability of primary and backup systems for augmenting the on-shift emergency response staff to determine the adequacy of licensee methods for staffing emergency response facilities in accordance with their emergency plan. The inspector reviewed the documents and references listed in the attachment to this report, to evaluate the licensees ability to staff the emergency response facilities in accordance with the licensees emergency plan and the requirements of 10 CFR Part 50, Appendix E. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one sample as defined in Inspection Procedure 71114.03-05.

b. Findings

No findings were identified.

1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies

Main article: IP 71114.05

a. Inspection Scope

The inspector reviewed summaries of corrective action program documents assigned to the emergency preparedness department and emergency response organization between August 1, 2008, and November 8, 2010, and selected 42 for detailed review against the program requirements. The inspector evaluated the response to the corrective action requests to determine the licensees ability to identify, evaluate, and correct problems in accordance with the licensee program requirements, planning standard 10 CFR 50.47(b)(14), and 10 CFR Part 50, Appendix E. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one sample as defined in Inspection Procedure 71114.05-05.

b. Findings

Introduction.

The inspector identified a violation of 10 CFR 50.47(b)(10) for the failure to develop and have in place guidelines for the choice of protective actions during an emergency consistent with federal guidance. This is a repeat issuance of this violation due to the failure to implement timely corrective actions. The violation was determined to be Green under the Significance Determination Process.

Description.

On July 30, 2009, during an operating experience review of a Green noncited violation issued to the Waterford Station Unit 3 (Inspection report 05000382/2009003) the licensee determined their guidelines for the choice of protective actions during an emergency were similar to those of Waterford, and were not consistent with federal guidance. Specifically, the licensees process for extending protective action recommendations as wind vectors change was not consistent with the guidance of EPA-400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents. The licensee identified their emergency response organization was trained to automatically extend existing protective action recommendations into all newly affected areas as the wind shifted without considering the radiation dose projected for those areas.

Procedure 10-S-01-12, Radiological Assessment and Protective Action Recommendation, Revision 40 contains the licensees guidelines for extending existing protective action recommendations. The licensees practices result in unnecessary recommendations for protective actions in areas where valid dose projections show federal protective action guides are not exceeded, and may expose members of the public to unjustified risks. This issue is documented in the licensees corrective action program as Condition Report CR-GGN-2009-3902 and CR-HQN-0757.

Shortly after the NRC conducted an inspection on September 8 through 11, 2009, and determined the licensee was committed to the EPA-400-R-92-001, protective action guides as a basis for recommending to offsite authorities protective actions for the public in the emergency planning zone. The inspector also determined that licensee practices

that automatically extended protective action recommendations as winds shifted did not comply with NRC requirements. This conclusion was documented as a licensee-identified noncited violation of 10 CFR 50.47(b)(10) in Inspection Report 05000416/2009004, issued October 29, 2009, (ADAMS ML093050016).

During this inspection, the inspector determined the licensee was still operating, training, and exercising using Procedure 10-S-01-12, Radiological Assessment and Protective Action Recommendation, Revision 40, that was in place on September 11, 2009. The inspector determined that the licensee had consulted with offsite authorities concerning the deficient condition, had reviewed corrective actions that had been planned and implemented for the same deficiency by other licensees, and had drafted a revision to Procedure 10-S-01-12, but had not implemented the procedure change.

To restore compliance with NRC requirements, the licensee implemented Procedure 10-S-01-12, Radiological Assessment and Protective Action Recommendation, Revision 41, on December 14, 2010. The inspector determined that this procedure revision did not require the emergency response organization to automatically recommend the extension of existing protective actions into newly-affected areas as wind direction changes, and did require the evaluation of radiation dose in newly-affected areas before changing protective action recommendations.

Analysis.

The failure to establish guidelines for the choice of protective actions during an emergency that were consistent with federal guidance was a performance deficiency.

This finding is more than minor because it affects the Emergency Preparedness Cornerstone objective of implementing adequate measures to protect the health and safety of the public during a radiological emergency, and is associated with the cornerstone attributes of emergency response organization performance and procedure quality. This finding was evaluated using the emergency preparedness significance determination process and was determined to be of very low safety significance (Green)because it was a failure to comply with NRC requirements, was associated with risk significant planning standard 50.47(b)(10) as defined in Inspection Manual Chapter 0609, Appendix B, Section 2.0, and was not a risk significant planning standard functional failure or a planning standard degraded function because appropriate protective action recommendations for the public would have been issued for all areas where protective action guides were exceeded. The finding is representative of current performance, is associated with a risk significant planning standard, and the licensee did not implement corrective actions for 504 days after the issue was identified. The finding is related to the corrective action element of the problem identification and resolution crosscutting aspect because the licensee failed to take corrective actions to address the safety issue in a timely manner P1.d].

Enforcement.

Title 10 CFR 50.47(b)(10) requires, in part, that guidelines for the licensees choice of protective actions during an emergency, consistent with federal guidance, are developed and in place. Federal guidance for the choice of protective actions during an emergency is described in EPA-400-R-92-001.

Contrary to the above, between July 30, 2009 and December 14, 2010, the licensee did not develop and have in place guidelines for the choice of protective actions during an

emergency that were consistent with federal guidance. The licensees guidelines for extending initial protective action recommendations under conditions of changing wind direction vectors were not consistent with EPA-400-R-92-001 guidance. Specifically, the licensees practice of automatically extending existing offsite protective action recommendations without evaluating dose assessment information was not consistent with federal guidance that evacuation was seldom justified for radiation doses below the protective action guides.

Inspection Manual Part 9900, Section 7.2, states, in part, that the NRC will consider safety significance, the effects on operability, the significance of the degradation, and what is necessary to implement corrective actions in determining whether the licensee is making reasonable efforts to complete corrective actions promptly. The guidance of Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Section 5.2, Timeliness, states, in part, that a risk significant planning standard-related drill/exercise performance weakness is typically corrected within 90 days of identification. Planning Standard 50.47(b)(10) is identified in Inspection Manual Chapter 0609, Appendix B, as a risk significant planning standard with substantial effect on the Emergency Preparedness Cornerstone objective. The actions necessary to implement corrective action included revising Procedure 10-S-01-12, Radiological Assessment and Protective Action Recommendation, and providing appropriate retraining to those emergency response organization personnel responsible for implementing the procedure. The licensee implemented Procedure 10-S-01-12, Revision 41, on December 14, 2010, restoring compliance with NRC requirements. The NRC concluded that 502 days was not a reasonable time for the licensee to restore compliance, considering the safety significance of the violation identified on July 30, 2009, and the actions necessary for the licensee to implement corrective actions.

The inspector sampled licensee submittals for the Drill and Exercise Performance, performance indicator for the period from the 3rd quarter 2009 through the 3rd quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, were used. The inspector reviewed the licensees records associated with the performance indicator to verify that the licensee accurately reported the indicator in accordance with relevant procedures and the Nuclear Energy Institute guidance. Specifically, the inspector reviewed licensee records and processes including

procedural guidance on assessing opportunities for the performance indicator; assessments of performance indicator opportunities during pre-designated control room simulator training sessions, performance during the 2009 biennial exercise, and performance during other drills. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of the drill/exercise performance sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.7 Emergency Response Organization Drill Participation (EP02)

a. Inspection Scope

The inspector sampled licensee submittals for the Emergency Response Organization Drill Participation performance indicator for the period from the 3rd quarter 2009 through the 3rd quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6, was used. The inspector reviewed the licensees records associated with the performance indicator to verify that the licensee accurately reported the indicator in accordance with relevant procedures and the Nuclear Energy Institute guidance. Specifically, the inspector reviewed licensee records and processes including procedural guidance on assessing opportunities for the performance indicator, rosters of personnel assigned to key emergency response organization positions, and exercise participation records. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of the emergency response organization drill participation sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

.8 Alert and Notification System (EP03)

a. Inspection Scope

The inspector sampled licensee submittals for the Alert and Notification System performance indicator for the period from the 3rd quarter 2009 through the 3rd quarter 2010. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspector reviewed the licensees records

associated with the performance indicator to verify that the licensee accurately reported the indicator in accordance with relevant procedures and the Nuclear Energy Institute guidance. Specifically, the inspector reviewed licensee records and processes including procedural guidance on assessing opportunities for the performance indicator and the results of periodic alert notification system operability tests. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of the alert and notification system sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

Main article: IP 71152

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

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

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

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

b. Findings

No findings were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

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

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

b. Findings

No findings were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensees corrective action program and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors focused their review on repetitive equipment issues, but also considered the results of daily corrective action item screening discussed in Section 4OA2.2, above, licensee trending efforts, and licensee human performance results. The inspectors nominally considered the 6-month period of June 1, 2010, through December 31, 2010, although some examples expanded beyond those dates where the scope of the trend warranted.

The inspectors also included issues documented outside the normal corrective action program in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance audit/surveillance reports, self-assessment reports, and maintenance rule assessments.

The inspectors compared and contrasted their results with the results contained in the licensees corrective action program trending reports. Corrective actions associated with a sample of the issues identified in the licensees trending reports were reviewed for adequacy.

These activities constitute completion of one semi-annual trend inspection sample as defined in Inspection Procedure 71152-05.

b. Findings and Observations

No findings were identified.

The inspectors identified an increasing trend in condition reports identifying poor maintenance practices on safety-related equipment. The specific items documented in the condition reports were reviewed by the inspectors and determined that all were

minor in nature. Several of the condition reports had been identified by NRC inspectors.

The licensee is aware of the adverse trend and is implementing a maintenance warranty program to address the deficiencies in maintenance work practices.

The inspectors specifically reviewed condition reports that documented equipment reliability issues and found no adverse trends, except for a higher number of issues related to components in the control room air conditioning system and the new mitigation monitoring system used for implementing online noble metal reactor coolant chemistry.

Using a keyword search, the inspectors reviewed adverse trends that had been previously identified in inspection report 05000416/2010003. The inspectors identified that several of these trends have not been resolved. The licensee is aware of these continuing adverse trends.

.4 Selected Issue Follow-up Inspection

a. Inspection Scope

During a review of items entered in the licensees corrective action program, the inspectors reviewed the disposition and corrective actions associated with condition report CR-GGN-2010-02674 which documented reactor level oscillations during the shutdown to refueling outage 17 (RF17). During the review, the inspectors found deficiencies in the classification, causal evaluation, and corrective actions associated with the event.

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

b. Findings

Introduction.

The inspectors identified a Green noncited violation of 10 CFR 50 Appendix B, Criterion V, for an inadequate shutdown procedure resulting in power and level oscillations in the reactor. The performance deficiency was self-revealing; however the inspectors added significant value by identifying inadequate condition report classification, causal evaluation, and corrective actions.

Description.

On April 25th, 2010, the station performed a shutdown of the reactor to enter the refueling outage (RF17). Four days prior to the start of the outage, station management chose to perform a soft shutdown of the reactor to promote faster cooling of the main turbine. This allowed turbine work to begin earlier in the outage, and would save about 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months of outage critical path time. The soft shutdown would require operators to reduce reactor power to approximately 18 percent prior to inserting a manual scram. Integrated Operating Instruction (IOI), 03-1-01-02, Power Operations was revised so that operators could reduce power below 25 percent power.

Subsequently, during the shutdown, with power at approximately 18 percent and a single feed pump in service with reactor coolant level control in single-element control, actual feedwater flow to the reactor began to decrease. As reactor level decreased, the single-element feedwater control system responded, increasing feedwater pump turbine speed

demand. However, the feed pump was initially unable to overcome reactor pressure to inject into the vessel, resulting in reactor water level decrease. The feedwater pump turbine eventually increased speed sufficiently to raise discharge pressure of the pump greater than reactor pressure, resulting in a reactor level increase. This cycle continued on a two minute frequency causing six oscillations in feed water level (level oscillations from +38.9 to +19.7 inches) and reactor power (four percent APRM power peak to peak)with increasing decay ratios. A turbine bypass valve opened when the turbine load control system automatically switched to bypass mode and was credited for stopping the oscillations. Ultimately, it was determined that the plant should have been placed in startup feedwater level control so that at the reduced feedwater flow rate, adequate feedwater pressure could be achieved to overcome reactor pressure.

Operations personnel originally documented this event in CR-GGN-2010-02674 as an unexpected condition caused by the turbine load control system, and the condition report was subsequently categorized as a C level which did not require a cause determination.

The condition report evaluation, performed by engineering staff, concluded that the turbine load control system responded appropriately and should have been expected by operations personnel. In addition, it was noted that the bypass valve opening did not initiate the level oscillations. The condition report also stated the following: There apparently were other factors occurring during the shutdown possibly related to single element control which affected water level. The evaluation was completed on May 2, 2010. The inspectors reviewed the closed condition report on June 23, 2010, and found that no new corrective action or condition report was initiated to address the unknown cause of the power and level transient. The inspectors brought this to the attention of station management, and the inspectors concern was documented in CR-GGN-2010-05140, which was categorized as a B level requiring a low-tier apparent cause evaluation. On November 30, 2010, the final cause evaluation was completed and concluded that the revision to the IOI failed to address the impacts to both the reactor water level control system and the turbine load control system.

Inspectors also questioned the adequacy of training performed prior to the outage. It is a station standard to practice critical evolutions in the simulator prior to operating the plant. In this case, the inspectors found that the just in time training did not use the revised power operations procedure, even though the procedure had been revised the day before. The inspectors also found that, contrary to standard industry practice, the licensee does not require testing of significant revisions to the integrated operating instructions in the simulator prior to their use.

Analysis.

Conducting a reactor shutdown with an inadequate procedure is a performance deficiency. This performance deficiency is more than minor and is therefore a finding because it was associated with the initiating events cornerstone attribute of procedure quality and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Using the Phase 1 Screening Worksheet in attachment 4 of Inspection Manual 0609, Significance Determination Process, the inspectors determined that the finding had very-low safety significance (Green) because in Table 4a of the Phase 1 Screening Worksheet, the finding was a transient initiator that

did not contribute to both the likelihood of a reactor trip and to the likelihood that mitigation equipment or functions would not be available. The inspectors determined that the apparent cause of this finding was that the licensee had failed to adopt a requirement that the proposed revision of IOI 03-1-01-02 was safe in order to proceed, and that this apparent cause is reflective of current licensee performance. This finding therefore has a cross-cutting aspect in the Decision-Making component of the Human Performance area because the licensee did not adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action H.1(b).

Enforcement.

10CFR50 Appendix B, Criterion V requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Contrary to this, on or about April 21, 2010, the licensee completed an activity affecting quality that was prescribed by a procedure that was not appropriate to the circumstances. Specifically, licensee operation of the reactor at power levels less than 25 percent power was prescribed by a procedure that did not properly account for the behaviour of various control systems at that power level. As a result, at power levels less than 25 percent power, licensee operation of the reactor in accordance with that procedure resulted in oscillations in reactor water level and reactor power. Because this violation was of very low safety significance and the licensee has entered it into their corrective action program as condition report CR-GGN-2010-05140, this violation is being treated as an NCV, consistent with Section 2.3 of the NRC Enforcement Policy, and has been designated NCV 05000416/2010005-04 Inadequate Reactor Shutdown Procedure Causes Power and Level Oscillations.

.5 Selected Issue Follow-up Inspection

a. Inspection Scope

During a review of items entered in the licensees corrective action program, the inspectors recognized CR-GGN-2010-01854, a corrective action item documenting a failure of the turbine first-stage pressure sensing line due to a steam leak in a flex hose.

The inspectors reviewed that item as described in Inspection Procedure 71152.02 to verify, in part, licensee evaluation and disposition of operability and reportability issues; consideration of extent of condition and cause, generic implications, common cause, and previous occurrences; classification and prioritization of the problems resolution commensurate with the safety significance; and identification of corrective actions that were appropriately focused to correct the problem.

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

b. Findings

No findings were identified.

.6 Selected Issue Follow-up Inspection

a. Inspection Scope

An inspection was performed at the Entergy corporate office in Jackson, Mississippi on June 14 through 17, 2010, to review the circumstances surrounding missed quality control (QC) verification inspections documented in CR-HQN-2009-01184 and CR-HQN-2010-00013. The issue involved QC verification inspections performed during construction-related activities which were required as part of the Entergy quality oversight and verification programs. The inspection was performed to determine if the licensee had taken corrective actions commensurate with the significance of the identified issues, and to assess the impact, if any, on the operability of plant equipment caused by the missed inspections. This inspection was conducted by inspectors from Regions I, II, and IV, as well as a Senior Program Engineer from the Quality and Vendor Branch of the Office of Nuclear Reactor Regulation (NRR). The inspection covered all NRC-licensed sites owned by Entergy Operations, Inc., including Arkansas Nuclear One, James A. Fitzpatrick, Grand Gulf Nuclear Station, Indian Point Units 2 and 3, Palisades Plant, Pilgrim Nuclear Power Station, River Bend Station, Vermont Yankee, and Waterford 3.

The inspectors reviewed root cause analyses documented in Condition Reports CR-HQN-2009-01184 and CR-HQN-2010-00013, and the results of the licensees extent of condition reviews and plant impact assessments. The inspectors also independently assessed the potential impacts of the missed inspections on the operability of plant equipment by reviewing all of the examples identified by the licensee, and by independently reviewing completed modifications and work orders to identify additional examples. The inspectors also reviewed the corrective action database to assess reported equipment failures in order to assess whether the failure might have involved missed QC verification inspections.

The inspectors assessed causal factors that may have contributed to missing QC verification inspections. This assessment included reviewing the Entergy Quality Assurance Program Manual (QAPM) requirements, changes made to the QAPM, and the level of agreement between the QAPM and its implementing procedures.

Specific documents reviewed are listed in the attachment.

b. Findings

Background The inspectors identified problems with the implementation of elements of the Quality Assurance (QA) Program that affected the fleet of Entergy Operations Inc., (hereafter referred to as Entergy) nuclear power plants that are licensed by the NRC. While the plant organizations are NRC licensees, Entergy also has corporate groups which are not NRC licensees that are actively involved in some activities affecting sites, including program and procedure changes. Entergy adopted a business strategy of adopting standard programs and procedures at all fleet plants.

On October 30, 2009, the NRC discussed with Entergy the initial concerns about whether QC verification inspections were being performed consistently for the types of work that require that level of inspection. Both the non-licensed and licensed Entergy organizations responded with an appropriate review of the issues. Entergys review of work documents that were potentially affected was extensive at each site. Entergys total review examined over 320 Engineering Change documents and 2676 Work Orders.

Of the 30 Work Orders identified to have QC verification inspection deficiencies affecting eight safety-related design changes, all 30 were determined by Entergy to have sufficient documentation to provide confidence that the equipment was installed correctly. Specific corrective actions were identified and implemented to ensure that QC verification inspections would be included in current and future work documents, including procedure enhancements.

The information provided to the NRC was used to perform a focused inspection in order to assess the impact of the missed verification inspections at each of the NRC-licensed facilities. The inspection documented below independently assessed the potential impact of missed QC verification inspections on the operability of plant equipment, as well as assessing details of QA Program for the Entergy fleet.

Two findings were identified during this inspection. These findings involved missed QC verification inspections at seven Entergy sites, and the assignment of individuals to the QA Manager position that did not meet the experience and qualification requirements at eight sites. Only the findings impacting this licensee are described below.

The inspectors concluded that the Entergy fleet organizational structure and Entergy strategy of adopting standardized procedures across the fleet were contributing factors to the findings. Specifically:

Changes to adopt the standard fleet QA program created a partial conflict with existing requirements for worker qualifications at some sites. The process for creating and revising standardized fleet procedures and programs used to meet NRC requirements must ensure that site-specific regulatory requirements and commitments are properly addressed for all sites.

Changes that removed details from existing site-specific QA and QC program implementing procedures while shifting to standardized fleet procedures contributed to the finding involving missed QC verification inspections. Condition reports at individual sites regarding problems related to this issue were not recognized collectively as symptoms of a problem with these procedures because they were addressed at the site level.

b.1 Failure to Perform Required Quality Control Inspections

Introduction.

The inspectors identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were included in quality-affecting procedures and work

instructions for construction-like work activities as required by the Quality Assurance Program.

Description.

In response to the inspectors request for information concerning implementation of the quality oversight and verification programs, the licensee performed a review of a representative sample of engineering changes and work order tasks issued between 2006 and 2009. The licensees review included performing equipment walkdowns, evaluating rework rates and human error rates, and causes for failures of significant components. Based on the results of these reviews, Entergy initiated condition reports at the various sites to document problems with Quality Control (QC) verification activities and failures to perform required QC reviews of safety-related engineering changes and construction related work activities. Entergys investigation concluded that procedures contained inadequate guidance, which resulted in inconsistent implementation of the QC Program. Specifically, some safety-related design change work orders were not reviewed to determine whether QC verification inspections were required, and some safety-related design change work orders did not include all required QC verification inspections. These examples were documented in CR-HQN-2009-01083, -01084, -01085, -01093, -01096, -01140, -01169, -01170, -

01184, and -01188.

Additional findings identified by Entergys review included:

Managers in maintenance organizations did not have a detailed understanding of QC responsibilities, required inspections, or what documents required review (CR HQN-2009-01150).

A weakness was identified in the process for ensuring proper approval of contract QC inspection personnel at all Entergy sites. Procedure EN-QV-111, Training and Certification of Inspection/verification and examination Personnel, Section 4.0 [1], required that the Manager responsible for Quality Assurance or designee at each location is responsible for approving ANSI N45.2.6 certification of QC inspection personnel. In practice, contract QC inspectors qualifications were not approved by the QA Manager prior to November 2009. This was determined to be a minor violation because the ANSI Level III inspector at each site was documenting that the contract QC personnel had the necessary qualifications to perform the inspections for which they were contracted. This issue was entered into the licensee's corrective action program as CR-HQN-2009-1091.

At individual Entergy plants, 27 condition reports were written in 2008 and 2009 to document potentially missed QC verification inspections or missed reviews to consider QC verification inspections prior to the NRC engaging Entergy on this issue. Of those, seven were actual missed inspections (CR-RBS-2009-05041, CR-JAF-2008-03648, and CR-PNP-2008-00916 and CR-PNP-2008-03922, CR-PNP-2009-01798, CR-PNP-2009-02059, and CR-PNP-2009-02255). Multiple condition reports documented work package quality issues that impacted the ability to identify appropriate QC verification inspection requirements.

Two examples of QC programmatic issues were identified, assigned the Entergy headquarters, and not properly addressed (CR-ANO-C-2009-01884, and CR-HQN-2009-00178). These were considered examples of the violation discussed below.

River Bend Station was using notification points instead of designating specific QC hold points (CR-RBS-2008-04685). This is further discussed in Section 4OA7.

Insufficient resources were assigned or qualified to perform the required tasks at Grand Gulf Nuclear Station and River Bend Station. River Bend Station operated with a single QC Level II inspector for more than 3 years, and Grand Gulf Nuclear Stations two QC inspectors did not have all of the discipline certifications for which they were conducting inspections (CR-HQN-2009-01140 and CR GGN-2009-06575). While these conditions were inappropriate, the inspectors did not identify a separate violation associated with these issues. To the extent that the individuals at River Bend Station were evaluating work documents for QC verification inspections and not correctly identifying those verifications, those examples are part of the violation discussed below.

Although equipment-related QC condition reports were addressed appropriately, QC programmatic issues were not always effectively addressed.

QA audits and oversight activities for the QC Program missed opportunities to identify the findings of their investigation (CR-HQN-2009-01169, CR-HQN-2009-0153, and CR-HQN-2010-00013). In particular, the Entergy corporate ANSI Level III inspector was required to perform periodic surveillances of QC inspection activities to ensure the program is being adequately implemented and maintained, but these required surveillances were not performed in 2008 (CR-HQN-2009-00111). This is further discussed in Section 4OA7.

Subsequent to the identification of these deficiencies, Entergy initiated prompt corrective actions to ensure that appropriate safety-related, engineering changes and non-routine maintenance work orders were identified and routed to the Maintenance Inspection Coordinator for evaluation and inclusion of QC verification inspections in accordance with the revised requirements of procedure EN-WM-105, Planning. These corrective actions and actions to preclude recurrence were collectively documented in the following Level A condition reports: CR-HQN 2009-01184, dated December 21, 2009 and CR-HQN-2010-0013, dated January 6, 2010.

In-office NRC reviews identified the need to conduct further inspection activities. On June 14 through 17, 2010, the inspectors conducted a focused review of work performed at each NRC-licensed Entergy site to assess whether examples of missed QC verification inspections identified by Entergy during their review had the potential to have impacted the operability of important plant equipment. The inspectors also reviewed the corrective action database and maintenance records to independently assess the rigor of the Entergy review and to identify additional examples of missed QC verification

inspections. The inspectors identified no additional examples, and concluded that the Entergy reviews were sufficient to identify the scope of the problems and develop actions to address the causes.

The inspectors reviewed specific work items whose scope met QAPM requirements to have had QC verification inspections but did not have the appropriate inspections.

Based in part on interviews with Entergy personnel, the inspectors determined that procedural guidance for work planning was not sufficiently detailed or clear to ensure that work packages with construction-like activities would be reviewed by the specified QC personnel. These individuals were responsible for designating the QC inspections that were required by the QAPM.

The inspectors also identified numerous condition reports written at Entergy sites that documented improper implementation of QC verification inspections. Specific condition reports are listed in the attachment.

In response to inspectors questions about staffing and qualifications among QC inspectors, the licensee identified that at Grand Gulf Nuclear Station, QC inspectors had been performing QC verification inspections in disciplines for which the individuals did not have certifications. This was addressed in CR-GGN-2009-06575 and CR-HQN-2009-01197. Relative to the QC staffing level at River Bend Station, the inspectors reviewed applicable regulatory requirements and concluded that, while this practice was inappropriate, it did not violate any specific requirement. Also, while QC inspector staffing was marginal at Grand Gulf Nuclear Station and River Bend Station, these stations frequently used contractors and inspectors from other Entergy sites to supplement the onsite QC inspection staff.

Analysis.

The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This programmatic deficiency, if left uncorrected, could lead to a more significant safety concern in that the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the Design Control attribute of the Mitigating Systems cornerstone because missed quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events.

This performance deficiency was determined to have very low safety significance in Phase 1 of Manual Chapter 0609, Significance Determination Process, since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. Specifically, inspectors verified by sampling that work documents provided objective quality evidence that work activities that had missed quality control verifications were properly performed.

The inspectors determined that this performance deficiency involved a cross-cutting aspect related to the Human performance in Decision-making (H.1(a)), because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether QC verification inspections were appropriate.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion X, Inspection, requires, in part, that: Examinations, measurements, or tests of materialshall be performed for each work operation where necessary to assure quality.If mandatory inspection hold points, which require witnessing or inspecting by the licensees designated representative and beyond which work shall not proceed without the consent of the designated representative are required, the specific hold points shall be indicated in appropriate documents.

Entergys QAPM, Revision 20, Section B.12., Inspection requires, in part, that:

Provisions to ensure inspection planning is properly accomplished are to be established. Planning activities are to identify the characteristics and activities to be inspected, the inspection techniques, the acceptance criteria, and the organization responsible for performing the inspection. Provisions to identify inspection hold points, beyond which work is not to proceed without consent of the inspection organization, are to be defined.

Contrary to the above, from February 2006, to December 2009, the licensee failed to ensure that examinations, measurements, or tests of material were performed for each work operation where necessary to assure quality, and failed to include mandatory inspection hold points in appropriate documents. Specifically, multiple examples of Maintenance Work Orders and Engineering Change documents for construction-related activities involving safety-related systems structures and components were identified where witnessing or inspections were required to be performed to ensure quality, but these steps were not identified, included in the work documents, or performed as required QC hold points in the work instructions. Condition reports documenting the specific problems and examples of the violation included:

CR-GGS-2009-06907 CR-GGS-2009-06920 CR-GGS-2009-06921 CR-GGS-2009-06922 CR-GGS-2009-06923 CR-GGS-2009-06927 CR-GGS-2009-06806 CR-HQN-2009-01083 CR-HQN-2009-01084 CR-HQN-2009-01085 CR-HQN-2009-01093

CR-HQN-2009-01096 CR-HQN-2009-01140 CR-HQN-2009-01169 CR-HQN-2009-01170 CR-HQN-2009-01184 CR-HQN-2009-01188 Because this issue was of very low safety significance and was entered into the corrective action program as Condition Reports CR-HQN 2009-01184 and CR-HQN-2010-0013, consistent with Section VI.A of the Enforcement Policy, this violation is being treated as a noncited violation, NCV 05000416/2010005-05 Failure to Perform Required Quality Control Inspections.

b.2. Failure to Implement the Experience and Qualification Requirements Associated With the Quality Assurance Program

Introduction.

The inspectors identified a Green noncited violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program.

Description.

During their review of the issues surrounding the improper implementation of quality control (QC) verifications discussed above, the inspectors noted that the root cause analysis documented in CR-HQN-2010-0013 identified that lack of experience of the Quality Assurance (QA) Manager contributed to the failure to identify the trend in missed QC verification inspections. The inspectors reviewed the relevant experience and qualifications of the QA Manager at each Entergy site. The inspectors also reviewed the NRCs safety evaluation report that approved Entergys original corporate Quality Assurance Program Manual (QAPM), which is the document that contains the QA Program. Additionally, the inspectors reviewed the administrative section of the Technical Specifications for all the Entergy sites and a sample of evaluations, performed in accordance with 10 CFR 50.54(a), that supported Entergy QAPM changes and alignment of plants that were subsequently purchased by Entergy.

The Entergy corporate QAPM required each site to meet the experience and qualification standards in ANSI/ANS 3.1-1978, American National Standard for Selection and Training of Nuclear Power Plant Personnel. Section 4.4 included qualification and experience requirements for the personnel described as group leaders of five professional-technical groups, including Quality Assurance. Section 4.4.5, Quality Assurance, required that the responsible person shall have six years experience in the field of quality assurance, preferably at an operating nuclear plant, or

operations supervisory experience. At least one year of this six years experience shall be nuclear power plant experience in the overall implementation of the quality assurance program. (This experience shall be obtained within the quality assurance organization.)

On December 15, 2008, procedure EN-QV-117, Oversight Training Program, the Entergy procedure used by all Entergy sites to implement the requirements of ANSI/ANS 3.1-1978, was revised by the Entergy corporate QA group. Section 5.7, Manager/QA Senior Auditor Training, was changed to state:

Either the QA Manager or the Senior QA Auditor will meet the requirements of ANS 3.1-1978 paragraph 4.4.5 for operating plants and if applicable ANS 3.1-1993 paragraph 4.3.7 for new plants.

The inspectors reviewed completed Personnel Change Planning Checklist/Forms for QA Managers at each site. Entergy used this form to evaluate QA Manager Candidates prior to the implementation of an Entergy fleet-wide restructuring in July 2007.

8, Change Management Guidelines for Alignment Implementation, included the following conclusion for the individual that subsequently was assigned to be the QA Manager:

[Individuals name redacted] meets the minimum requirements for QA Manager with the exception of at least one year of this six years experience shall be nuclear power plant experience in the overall implementation of the quality assurance program. This requirement must be met by the QA Senior Auditor.

Based on discussions with Entergy corporate QA personnel, the inspectors determined that Entergy personnel had interpreted ANSI/ANS 3.1-1978, Sections 4.4 and 4.4.5 to allow the Senior Auditor to be considered the QA group leader described in the standard for purposes of meeting the experience requirements of Section 4.4.5 in cases where a candidate for the position of QA Manager did not satisfy the experience requirements.

In reviewing this issue, the NRC staff has determined that the group leader in this case is the individual filling the position assigned responsibility for overall implementation of the QA Program (Entergy used the title QA Manager for this position). The individual meeting the experience and qualification requirements must be the individual assigned the responsibilities for overall implementation of the QA Program assigned within the QA Program.

The inspectors determined that this change to procedure EN-QV-117 did not ensure that the qualifications for the QA Manager would meet the requirements of standard. The inspectors identified an example where the Senior Auditor was credited as being the group leader for purposes of meeting ANSI/ANS 3.1-1978, and the individual who was assigned as the QA Manager did not meet the ANSI/ANS 3.1-1978 experience requirements. The team also determined that the responsibilities assigned to the QA Manager under the QAPM were not reassigned to the Senior Auditor, and the Senior Auditor did not report directly to the designated senior executive. The Senior Auditor

continued to report to the QA Manager, so the person with the greater experience did not have the positional authority to decide issues.

Analysis.

Failure to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This performance deficiency was determined to be more than minor because, if left uncorrected, it could create a more significant safety concern. Failure to have a fully qualified individual providing overall oversight to the QA Program had the potential to affect all cornerstones, but this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative significance determination, so it was assessed using IMC 0609, Appendix M, so it was evaluated using the qualitative criteria listed in Table 4.1. This finding was determined to be of very low safety significance because other quality assurance program functions remained unaffected by this performance deficiency, so defense-in-depth continued to exist. The inspectors determined that there was no cross-cutting aspect associated with this finding because this issue was not indicative of current performance because the violation occurred more than 3 years ago.

Enforcement.

Appendix B to 10 CFR 50, Criterion II, Quality Assurance Program, requires, in part, that the licensee establish a quality assurance program which complies with Appendix B. This program shall be documented by written policies, procedures, or instructions and shall be carried out throughout plant life in accordance with those policies, procedures, or instructions. The program shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained.

The Entergy Quality Assurance Program Manual, Revision 13, is the document used at each Entergy-owned site to describe the quality assurance program. Table 1, Section A of the Quality Assurance Program Manual states, in part, that qualifications and experience for station personnel shall meet ANSI/ANS 3.1-1978 except for positions where an exception to either ANSI/ANS 3.1-1978 or N18.1-1971 is stated in the applicable units Technical Specifications.

ANSI/ANS 3.1-1978, Section 4.4.5, Quality Assurance, states, in part, that the responsible person (i.e. the Quality Assurance Manager) shall have six years experience in the field of quality assurance. At least one year of this six years experience shall be obtained within the quality assurance organization.

Contrary to the above, between July 7, 2007, and July 8, 2008, the licensee failed to implement the quality assurance program requirements intended to provide indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency was achieved and maintained. Specifically, the individual(s) assigned to be the responsible person for the licensees overall implementation of the Quality Assurance Program did not have at least 1 year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. Because this

issue was of very low safety significance and was entered into the corrective action program as Condition Report CR-HQN-2010-00386, consistent with Section VI.A of the Enforcement Policy, this violation is being treated as a noncited violation, NCV 05000416/2010005-06: Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program.

.7 In-depth Review of Operator Workarounds

a. Inspection Scope

The inspectors reviewed the licensees identified Operator Workarounds (OWAs) and Operator Burdens (OBs) to verify the licensee is effectively implementing the actions of the OWAs and OBs, as well as taking appropriate corrective actions to correct the degraded or non-conforming conditions. The inspectors also interviewed licensed and non-licensed operators and performed frequent tours of the control room and the plant to verify that OWAs are being identified at an appropriate threshold and entered into the corrective action program.

b. Findings

No findings were identified.

4OA3 Event Follow-up

Main article: IP 71153

.1 Steam Leak in the Containment

a. Inspection Scope

On November 8, 2010, the inspectors responded to the control room to observe operator response to a steam leak in containment. The newly installed mitigation monitoring system positive displacement pump ejected the cylinder causing an approximate seven gallons per minute reactor coolant leak. The inspectors observed operator actions, control room briefs and overall plant response to the event. The inspectors also observed control room indications that control room staff used to identify abnormal conditions in the containment building. Documents reviewed for this inspection are listed in the attachment.

b. Findings

No findings were identified

.2 Steam Leak in the B High Pressure Heater Room

a. On December 15, 2010, the inspectors responded to the site and observed actions being taken to stop a steam leak in the B high pressure feedwater room. The leak was located on the welded pipe fitting connection on the 24-inch extraction steam piping at the high point vent 3/4-inch line. The inspectors monitored the licensees actions to stop the steam leak including attending briefings, reviewing engineering change packages to

perform a repair and interfaced with licensee management to determine their plan of action. The licensee reduced power to approximately 81 percent on December 24, 2010, and entered the room and placed an enclosed structure and stopped the leak.

Documents reviewed for this inspection are listed in the attachment.

b. Findings

No findings were identified

4OA6 Meetings

Exit Meeting Summary

On November 2, 2010, the inspector discussed the results of the licensed operator annual requalification examination with Mr. Richard Sumrall, Superintendent, Operations Training.

A telephonic exit meeting was held with Mr. Sumrall on November 3, 2010. The licensee acknowledged the results of the inspection presented in the final exit meeting. The inspector confirmed that proprietary information was not provided or examined during the inspection.

On November 19, 2010, the inspector presented the onsite emergency preparedness inspection results to Mr. R. Douet, Site Vice President, and other members of the licensees staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On January 10, 2011, the inspector presented the results of the Selected Issue Follow-up Inspection of quality assurance and quality control issues to Mr. M. Richey, Director, Nuclear Safety Assurance, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On January 11, 2011, the inspectors presented the inspection results to Jeremy Browning, General Plant Manager, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section 2.3 of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a noncited violation.

1. Title 10 CFR 50.47(b)(4) requires, in part, that a standard classification and emergency action level scheme is in use by the licensee. Contrary to the above, on October 21, 2010, the licensee identified they had not maintained in effect a standard classification and emergency action level scheme. Specifically, the licensee identified 394 examples of failures to implement emergency action level AA1-1 for airborne and liquid effluent

releases from October 1, 2008, to October 21, 2010. In each example, the 400-times discharge permit-specific alarm setpoint value could not be read on the associated process radiation monitor because the value was above the monitors operating range.

This finding was more than minor because it impacted the Emergency Preparedness Cornerstone objective attribute of emergency response organization performance and was evaluated as having very low safety significance (Green) because it was a failure to comply with NRC requirements, was associated with a risk-significant planning standard, and was not a functional failure or degraded function of the planning standard. This condition was documented in Condition Report CR-GGN-2010-07456. Because the finding was determined to be of very low safety significance and was entered into the corrective action program, this violation is being treated as a Green noncited violation consistent with the NRC Enforcement Policy.

2. With one or more fire door(s) being inoperable and with at least one side of the

inoperable fire door(s) having verified continuous fire monitoring and alarm capability, Condition 2.C.41 of Grand Gulf Nuclear Power Stations Facility Operating License requires the establishment of an hourly fire watch patrol in accordance with the station fire protection program. Contrary to the above, on September 19, 2010, several inoperable fire doors were not inspected by an hourly fire watch patrol for a period of approximately 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . This issue was documented in the licensees corrective action program in condition report CR GGN-2010-06824. This finding is of very low safety significance because at least one side of each inoperable fire door had verified continuous fire monitoring and alarm capability.

3. Procedure, EN-QV-111, Training and Certification of Inspection/Verification and

Examination Personnel, Section 4.0 [4](i), requires that the Entergy corporate ANSI Level III inspector shall perform periodic (annual) surveillances of quality control inspection activities to ensure that the program is being adequately implemented and maintained. Contrary to the above, no surveillances of quality control inspection activities were performed for any Entergy site during calendar year 2008. The issue was not suitable for quantitative significance determination, so it was assessed using IMC 0609, Appendix M, so it was evaluated using the qualitative criteria listed in Table 4.1.

This finding was determined to be of very low safety significance because other quality assurance program functions remained unaffected by this performance deficiency, so defense-in-depth continued to exist. This issue was entered into the licensee's corrective action program as CR-HQN-2009-00111.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

J. Abisamra, Echelon Chief Engineer

S. Beagles, Echelon Manager of Fleet Operations

J. Browning, General Plant Manager

R. Byrd, Echelon Sr. Staff Engineer

D. Coulter, Senior Licensing Specialist

J. Dent, Echelon General Manager Plant Operations, Fleet Operations Support

R. Douet, Site Vice President

B. Ford, Echelon Sr. Manager, Nuclear Safety and Licensing

G. Giles, Corrective Action Program Manager

E. Harris, Echelon, QA Manager

K. Higgenbotham, Operations Manager

J. Houston, Maintenance Manager

R. Jackson, Licensing

D. Jacobs, Echelon Sr. Vice President of Planning, Development and Oversight

J. Keir, GGNS Quality Specialist

C. Lewis, Manager, Emergency Preparedness

J. McCann, White Plains Vice President of Nuclear safety, Emergency Preparedness, and

Licensing

P. Morris, Echelon Manager of Administrative Services

T. Palmisano, Echelon Vice President of Oversight

C. Perino, Licensing Manager

R. Pownall, GGNS Sr. QA Auditor

W. Renz, Corporate Director, Emergency Planning

M. Richey, Director, Nuclear Safety Assurance

F. Rosser, Supervisor, Dosimetry

R. Sumrall, Superintendant, Operations Training

T. Tankersly, Echelon Director of Oversight

T. Trichell, Radiation Protection Manager

D. Tucker, Senior Emergency Planner

R. VanDenAkker, Senior Emergency Planner

E. Weinkam, White Plains Sr. Manager of Nuclear Safety and Licensing

D. Wiles, Engineering Director

R. Wilson, Manager, Quality Assurance

NRC Personnel

M. Ashley, Office of Nuclear Reactor Regulation

P. Elkmann, Senior Emergency Preparedness Inspector

K. Fuller, Region IV

M. Gray, Region I

J. Geissner, Region III

M. Haire, Region IV

N. Hilton, Office of Enforcement

D. Holody, Region I

A- 1 - Attachment

D. Jackson, Region I

W. Jones, Region IV

R. Kellar, Region IV

M. Marsh, Office of General Counsel

M. McLaughlin, Region I

M. Murphy, Office of Nuclear Reactor Regulation

C. Schulten, Office of Nuclear reactor Regulation

D. Thatcher, Office of Nuclear Reactor Regulation

A- 2 - Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

Inadequate Operability Evaluation Following a Spurious

05000416/2010005-01 NCV Actuation of the Standby Service Water Pump House Ventilation Fan (Section 1R15)

Untimely Corrective Actions in Response to Deficiencies in the 5000416/2010005-02 NCV RCIC Flow Control System (Section 1R22)

Failure to Have Guidelines for the Choice of Protective Actions

05000416/2010005-03 VIO During an Emergency Consistent with Federal Guidance (Section 1EP5)

Inadequate Reactor Shutdown Procedure Causes Power and

05000416/2010005-04 NCV Level Oscillations (Section 4OA2)

Failure to Perform Required Quality Control Inspections

05000416/2010005-05 NCV (Section 4OA2)

Failure to Implement the Experience and Qualification

05000416/2010005-06 NCV Requirements of the Quality Assurance Program (Section 4OA2)

A- 3 - Attachment

LIST OF DOCUMENTS REVIEWED