IR 05000271/2007004

IR 05000271-07-004, on 07/01/07 - 09/30/07, Vermont Yankee Nuclear Power Station; Maintenance Effectiveness

ML073110213
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	11/07/2007
From:	Racquel Powell NRC/RGN-I/DRP/PB5
To:	Ted Sullivan Entergy Nuclear Operations
Powell R, RI/DRP/610-337-6967
References
IR-07-004
Download: ML073110213 (47)

Text

UNITED STATES ber 7, 2007

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000271/2007004

Dear Mr. Sullivan:

On September 30, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Vermont Yankee Nuclear Power Station. The enclosed report documents the inspection results which were discussed on October 9, 2007, with you and members of your staff.

This inspection 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.

This report documents two self-revealing findings of very low safety significance (Green). One of these findings was determined to involve a violation of NRC requirements. In addition, a licensee-identified violation of very low safety significance is listed in this report. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest either of the NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Vermont Yankee Nuclear Power Station.

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

Sincerely,

/RA by Ronald R. Bellamy For/

Raymond J. Powell, Chief Projects Branch 5 Division of Reactor Projects Docket No.: 50-271 License No.: DPR-28 Enclosure: Inspection Report 05000271/2007004 w/Attachment: Supplemental Information cc w/encl:

M. Kansler, President, Entergy Nuclear Operations, Inc.

G. Taylor, Chief Executive Officer, Entergy Operations J. Wayne Leonard, Chairman and Chief Executive Officer, Entergy Operations J. Herron, Senior Vice President M. Balduzzi, Senior Vice President, Northeastern Regional Operations J. DeRoy, Vice President, Operations Support O. Limpias, Vice President, Engineering Sr Vice President, Engineering and Technical Services D. Mannai, Manager, Licensing, Vermont Yankee Nuclear Power Station Operating Experience Coordinator, Vermont Yankee Nuclear Power Station W. Maguire, General Manager, Plant Operations, Entergy Nuclear Operations, Inc.

N. Rademacher, Director, Engineering, Vermont Yankee Nuclear Power Station J. McCann, Director, Licensing Manager, Licensing E. Harkness, Director of Oversight, Entergy Nuclear Operations, Inc.

Assistant General Counsel, Entergy Nuclear Operations, Inc.

J. Sniezek, PWR SRC Consultant M. Lyster, PWR SRC Consultant G. Edwards S. Lousteau, Treasury Department, Entergy Services, Inc.

D. O Dowd, Administrator, Radiological Health Section, DPHS, DHHS, State of New Hampshire W. Irwin, Chief, CHP, Radiological Health, Vermont Department of Health Chief, Safety Unit, Office of the Attorney General, Commonwealth of Mass.

D. Lewis, Pillsbury, Winthrop, Shaw, Pittman LLP G. Bisbee, Esquire, Deputy Attorney General, Environmental Protection Bureau J. Block, Esquire J. Matteau, Executive Director, Windham Regional Commission D. Katz, Citizens Awareness Network (CAN)

R. Shadis, New England Coalition Staff G. Sachs, President/Staff Person, c/o Stopthesale J. Volz, Chairman, Public Service Board, State of Vermont

SUMMARY OF FINDINGS

IR 05000271/2007004; 07/01/07 - 09/30/07; Vermont Yankee Nuclear Power Station;

Maintenance Effectiveness.

This report covered a 13-week period of inspection by resident and region-based inspectors.

Two Green self-revealing findings, one of which was a non-cited violation (NCV), were identified. The significance of most findings is indicated by their color (Green, White, Yellow,

Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP 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.

Cornerstone: Initiating Events

Green.

A self-revealing NCV of TS 6.4, Procedures, was identified for Entergys failure to effectively incorporate readily available industry operating experience (OE) into the cooling tower (CT) inspection program and processes. Specifically, Entergy had not recognized the importance of performing hands-on inspections of CT structural members which were located in heavily loaded and normally inaccessible areas for detecting degraded conditions. As a result of not performing adequate inspections of B and C columns within the fill area, a partial collapse of non-safety CT cell 2-4 occurred on August 21, 2007. This event resulted in Entergy rapidly reducing reactor power from 100 percent to approximately 35 percent power, although no significant degradation of the safety-related cell was identified.

The performance deficiency is that Entergy did not incorporate readily available OE into the CT inspection process and procedures to detect degraded structural components.

This finding is more than minor because it is associated with the Equipment Performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors conducted a Phase 1 screening in accordance with IMC 0609, Appendix A. The finding was determined to be of very low safety significance because it did not contribute to both the likelihood of a reactor scram and the likelihood that mitigating equipment or functions would not be available. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Entergy did not implement and institutionalize OE through changes to station processes and procedures for the CT, as appropriate. P.2(b) (Section 1R12.1)iii

Summary of Findings (cont.)

Green.

A self-revealing Finding of very low safety significance was identified for Entergys failure to specify adequate preventive maintenance (PM) for main turbine stop valve number two (TSV-2). As a result, during troubleshooting activities on the TSV-2 bypass control mechanism, a reactor scram occurred when all four main turbine stop valves closed.

The performance deficiency is the failure of Entergy to have an adequate PM strategy for TSV-2. This finding is more than minor because it is associated with the Equipment Performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, during troubleshooting of the TSV-2 bypass control mechanism, an automatic reactor scram occurred. The inspectors conducted a Phase 1 screening of the finding in accordance with IMC 0609, Appendix A. The finding was determined to be of very low safety significance because it did not contribute to both the likelihood of a reactor scram and the likelihood that mitigating equipment or functions would not be available. The finding has a cross-cutting aspect in the area of Human Performance, Resources component, because Entergy did not maintain an effective PM program for TSV-2.

Specifically, the PM for TSV-2 did not specify a periodic activity to inspect, rebuild, and lubricate the bell crank assembly portion of the bypass control mechanism. H.2(c)

(Section 1R12.2)

B. Licensee-Identified Findings One violation of very low safety significance (Green), which was identified by the licensee, has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and corrective actions are listed in Section 4OA7 of this report.

iv

REPORT DETAILS

Summary of Plant Status

Vermont Yankee (VY) Nuclear Power Station began the inspection period operating at 100 percent power. On August 4, 2007, power automatically decreased to approximately 80 percent when the C reactor feed pump unexpectedly tripped. Entergy determined that the feed pump tripped as a result of blockage in the systems feed flow sensing line. Following repairs, operators returned the C feed pump to service and increased reactor power to 100 percent on August 5.

On August 21, operators reduced power to approximately 35 percent in response to a partial collapse of the non-safety related, forced draft, west cooling tower (CT) cell 2-4. Following the event, operators secured circulating water to the west CT to investigate and repair. From August 21 through August 30, power was maintained between approximately 35 to 65 percent.

On August 30, an automatic reactor scram occurred from approximately 63 percent power after all four main turbine stop valves unexpectedly closed during stop valve maintenance activities.

On September 1, the reactor was restarted and placed on the grid after the cause of the scram had been identified and corrected. Operators increased power to approximately 50 percent on September 2. From September 2 until September 15, power was maintained between approximately 50 to 70 percent.

On September 15, the west CT was placed back in service. Power was subsequently increased to 100 percent on September 16. On September 16, a downpower to approximately 70 percent was performed for a planned control rod pattern adjustment. On September 17, the plant was restored to approximately 100 percent power, where it remained for the rest of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

a.

Inspection Scope (1 sample)

The inspectors reviewed Entergys measures to ensure the readiness of the circulating water, service water (SW), and alternate cooling (ACS) systems for summer weather.

The intent of the inspection was to ensure adequate cooling remained available for both the non-safety related main condenser and for the safety-related SW and ACS systems during peak summer temperatures. The inspectors performed walkdowns of accessible portions of the cooling water systems to determine if they were properly aligned and maintained, in accordance with station procedures. The inspectors reviewed heat sink temperature and main condenser back pressure data to determine if adequate cooling was available. In addition, the inspectors reviewed condition reports (CRs) to determine if Entergy had properly identified and addressed known deficiencies. A listing of documents reviewed is provided in the Attachment to this report.

b. Findings

No findings of significance were identified.

1R02 Evaluations of Changes, Tests, or Experiments

a. Inspection Scope

(7 evaluations and 15 applicability determinations)

The inspectors reviewed safety evaluations to determine if changes and tests were evaluated and documented in accordance with 10 CFR 50.59; and, if required, Entergy obtained NRC approval prior to implementation. The inspectors assessed the adequacy of the safety evaluations through interviews with Entergy personnel and review of supporting information, such as calculations, engineering analyses, design change documentation, the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications (TS). In addition, the inspectors reviewed the administrative procedures that control the screening, preparation, and issuance of the safety evaluations to evaluate whether the procedures adequately implemented the requirements of 10 CFR 50.59, Changes, Tests, and Experiments. The inspectors also reviewed changes that Entergy had evaluated, using their applicability determination process, and determined that safety evaluations were not required. The inspectors performed this review to assess Entergys conclusions with respect to 10 CFR 50.59 applicability. The safety evaluations, applicability determinations, and screenings were selected based on the safety significance of the affected structures, systems, and components (SSCs).

The inspectors also reviewed issues that had been entered into the corrective action program to determine whether Entergy had been effective in identifying and resolving problems associated with the 10 CFR 50.59 safety evaluation process. A listing of the safety evaluations, safety evaluation screenings, applicability determinations, and other documents reviewed is provided in the Attachment to this report.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Partial Equipment Alignment

a. Inspection Scope

(3 samples)

The inspectors performed partial system walkdowns of the following risk-significant systems to evaluate the system alignment and to identify any discrepancies that could impact system operability. Observed plant conditions were compared to the standby alignment of equipment specified in applicable Entergy operating procedures (OP) and piping and instrumentation drawings (P&IDs). The inspectors observed valve positions, the availability of power supplies, and the general condition of selected components. A listing of documents reviewed is provided in the Attachment of this report.

• Diesel-driven fire pump while the C SW pump was out of service for emergent repairs;

• A Residual Heat Removal (RHR) and A RHR Service Water (RHRSW) systems while the B RHR was out of service for planned maintenance; and

• SW system while the ACS was inoperable with the deep basin drained.

b. Findings

No findings of significance were identified.

.2 Complete Equipment Alignment

a. Inspection Scope

(1 sample)

The inspectors performed a complete equipment alignment inspection of the accessible portions of the SW system. The inspectors compared the actual system configuration to approved P&IDs, the UFSAR, the SW system design basis document (DBD), OPs, and vendor manual requirements. The inspectors reviewed required support systems and ancillary equipment to determine if they interfered with the operation of system valves and if pump oil levels were maintained in the nominal band, and whether deficiencies had been entered into the corrective action program. The inspectors evaluated whether major system components were properly ventilated, and whether hangers and supports were correctly installed and functional. In addition, the inspectors evaluated a sample of previously identified deficiencies to determine if they had been properly addressed, and whether open items impacted system operability. A listing of documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

(12 samples)

The inspectors selected fire areas (FAs) important to plant risk, based on a review of Entergys Safe Shutdown Capability Analysis, the Fire Hazards Analysis, and the Individual Plant Examination External Events (IPEEE). The inspectors toured these areas to determine the suitability of Entergys control of transient combustibles and ignition sources, and the material condition and operational status of fire protection systems, equipment, and barriers. A listing of documents reviewed is provided in the of this report. The following FAs and fire zones (FZs) were inspected:

• High pressure coolant injection (HPCI) corner room (FZ RB1);

• Reactor core isolation cooling (RCIC) corner room, 232 foot elevation (FZ RB1S);

• B emergency core cooling system (ECCS) corner room (FZ RB2);

• Reactor building, 318 foot elevation (FZ RB7);

• Reactor building, 345 foot elevation (FZ RB7);

• Circulating water pump room (FZ 14);

• Service water pump room (FZ 15);

• West cooling tower (FA 16);

• RCIC corner room, 213 foot elevation (FA RCIC);

• Discharge structure (No fire designation);

• Alternate offgas (AOG) building (No fire designation); and

• East Cooling Tower (No fire designation).

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

a. Inspection Scope

(1 sample)

The inspectors reviewed Entergys established flood protection barriers and procedures for coping with internal flooding in the Reactor Building 318 foot elevation. The inspectors reviewed internal flooding design information associated with this location in Entergys IPEEE, the UFSAR, and in the Internal Flooding DBD. The inspectors conducted a walkdown of the area to determine if equipment and structures needed to mitigate an internal flooding event were available, as described in Entergy design documents. A listing of documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

a. Inspection Scope

(1 sample)

The inspectors observed a simulator-based licensed operator requalification annual exam on August 8, 2007. The inspectors evaluated crew performance in the areas of clarity and formality of communications; ability to take timely actions; prioritization, interpretation, and verification of alarms; procedure usage; control board manipulations; and command and control. Crew performance in these areas was compared to Entergy management expectations and guidelines as presented in Administrative Procedure (AP) 0151, Responsibilities and Authorities of Operations Department Personnel, AP 0153, Operations Department Communication and Log Maintenance, and Department Procedure (DP) 0166, Operations Department Standards. The inspectors also compared the simulator configuration with the actual control board configuration.

Finally, the inspectors observed the Entergy evaluators discuss identified weaknesses with the crew and/or individual crew members, as appropriate. A listing of documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

(3 samples)

The inspectors reviewed Entergy's evaluation of degraded conditions, involving safety-related SSCs for maintenance effectiveness during this inspection period. The inspectors reviewed Entergys implementation of the Maintenance Rule, 10 CFR 50.65, to determine if the conditions associated with the referenced CRs were appropriately evaluated against applicable Maintenance Rule functional failure criteria, as found in Entergy scoping documents and procedures. The inspectors discussed these issues with the system engineers and Maintenance Rule coordinators to determine if they were appropriately tracked against each system's performance criteria and that the systems were appropriately classified in accordance with Maintenance Rule implementation guidance. Documents reviewed during the inspection are listed in the Attachment. The following issues and/or systems reviewed were:

C RCIC outboard steam isolation valve, V13-16, failed local leakage rate test; C Partial collapse of CT cell 2-4; and C Reactor scram during #2 main turbine stop valve troubleshooting.

.1 Inadequate Inspection Program Resulted in a Partial Collapse of Cooling Tower Cell 2-4

Introduction:

A self-revealing non-cited violation (NCV) of TS 6.4, Procedures, was identified for Entergys failure to effectively incorporate readily available industry operating experience (OE) into the CT inspection program and processes. Specifically, Entergy had not recognized the importance of performing hands-on inspections of CT structural members which were located in heavily loaded and normally inaccessible areas for detecting degraded conditions. As a result of not performing adequate inspections of B and C columns within the fill area, a partial collapse of non-safety CT cell 2-4 occurred on August 21, 2007. This event resulted in Entergy rapidly reducing reactor power from 100 percent to approximately 35 percent power.

Description:

On August 21, 2007, a portion of CT cell 2-4 collapsed while the plant was operating at full power. To maintain State thermal discharge requirements, Entergy reduced power from 100 percent to approximately 35 percent. The inspectors assessment of Entergys performance in response to the event is in Section 4OA3.3 of this report.

There are two cooling towers at Vermont Yankee; each tower contains 11 cells. The non-safety function of the cooling towers is to reduce the temperature of the water used to cool the plant before it is returned to the Connecticut River. Only one cell (cell 2-1) is safety-related. The safety-related function of cell 2-1 is to provide an alternate means to remove reactor heat following a plant shutdown in the unlikely event that the normal heat removal system becomes unavailable. Adjacent cell 2-2 is not safety-related, but is structurally robust and is designed to physically separate and protect cell 2-1.

Entergy completed a root cause analysis (RCA) to identify the causes associated with the partial cell collapse. Entergy determined that root causes included both a mechanistic and programmatic aspect:

• Mechanistic Aspect - Entergy concluded that a number of the wooden 4x4 B vertical columns located inside the fill area of cell 2-4 had failed prior to the cell collapsing. The columns were already heavily loaded from the weight of the water distribution deck and the circulating water header. The columns were further weakened due to stresses from a chemical iron-salt attack related to iron bolting used to connect the wooden columns, a biological fungal attack in areas that the wooden columns had been affected by the iron-salt, and from over-tightened bolts at spliced locations. The column failures caused the distribution deck to sag and the water header to separate; this resulted in additional water on the distribution deck, which increased the loading on the support columns. The added weight on columns that were already stressed caused additional columns to buckle, resulting in the collapse of a portion of the cell.

• Programmatic Aspect - The CT inspection program did not require inspections of the B and C columns in the normally inaccessible fill area. Specifically, Entergy had routinely performed remote boroscopic and/or visual inspections, but had not recognized the importance of utilizing hands-on inspections techniques to detect degraded structural conditions, such as iron-salt and fungal attack or over-tightened bolts at spliced locations.

Entergys RCA team reviewed internal and external OE reports related to CT structural events, which shared a common theme related to the importance of performing detailed inspections in the normally inaccessible fill areas. Additionally, the OE stated that traditional inspection methods of looking for signs of wood distress were not sufficient to identify degraded columns. Entergy identified that, although the OE had been received from offsite sources and Nuclear Electric Insurance Limited reports, no actions had been developed for Entergys CT inspection programs. The inspectors determined that multiple opportunities existed for Entergy to identify and incorporate the recommendations of the external and internal OE into the CT inspection process.

The performance deficiency is that Entergy did not incorporate readily available OE into the CT inspection process and procedures to detect degraded structural components.

Analysis:

This finding is more than minor because it is associated with the Equipment Performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors conducted a Phase 1 screening of the finding in accordance with Inspection Manual Chapter (IMC) 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The finding was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor scram and the likelihood that mitigating equipment or functions would not be available.

This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Entergy did not implement and institutionalize OE through changes to station processes and procedures for the CTs, as appropriate. P.2(b)

Enforcement:

TS 6.4, Procedures, requires, in part, that written procedures be established, implemented, and maintained for preventive and corrective maintenance operations which could have an effect on the safety of the reactor. Although no significant degradation of the safety-related cell was identified, there are substantive differences between the construction and maintenance of the safety-related and non-safety related cells, aspects of the inspection inadequacies associated with cell 2-4 are applicable to safety-related cell 2-1. As such, contrary to the above, work orders (WOs) which implemented vendor manual VY-EM-0146, Cooling Tower - Operation, Maintenance, Repair, and Inspections CT 2-1, did not incorporate readily available OE into the CT inspections to detect degraded structural components (i.e., the WOs did not specify periodic fill removal and/or hands on inspection of timbers). Corrective actions, taken or planned, included a physical inspection of both CTs, focusing on B and C columns, and the repair of identified structural components; the incorporation of OE into the procedures for future inspections; ensuring all identified cooling tower deficiencies are documented into the corrective action program; and completion of a corrective action effectiveness review within one year. Because this issue is of very low safety significance (Green) and has been entered into Entergys corrective action program (CR-2007-03243), this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy. NCV 05000271/2007004-01, Inadequate Inspection Program Resulted in the Partial Collapse of a Non-Safety-Related Cooling Tower Cell

.2 Reactor Scram During Main Turbine Stop Valve Troubleshooting

Introduction:

A self-revealing Finding of very low safety significance (Green) was identified for Entergys failure to specify adequate preventive maintenance (PM) for main turbine stop valve number two (TSV-2). As a result, during troubleshooting activities on the TSV-2 bypass control mechanism, a reactor scram occurred when all four main turbine stop valves closed.

Description:

On August 30, 2007, Entergy and vendor personnel were troubleshooting the failure of TSV-2 to re-open during TSV surveillance testing. During the troubleshooting, the bell crank assembly portion of the bypass control mechanism was observed to be sticking and not allowing TSV-2 to re-open. When the linkage was depressed, the assembly that had been sticking was suddenly freed, causing the TSV to open faster than normal. This caused a hydraulic pressure transient condition in the emergency trip oil system, which caused all four TSVs to close, and resulted in a reactor scram. The plant responded to the TSV closure and reactor scram as designed. An assessment of operator performance in response to the event is documented in Section 4OA3.4 of this report.

The function of the TSVs is to close on a turbine trip condition. The valves also provide a signal to the reactor protection system if 3 or more of the TSVs are less than 90 percent open with the reactor greater than 25 percent power, as occurred during this event. The design of TSV-2 is different from the other three TSVs. TSV-2 has an internal bypass valve which allows for turbine warmup, and an associated bypass control mechanism.

Entergy determined the root cause of the event to be the lack of a PM activity to inspect, rebuild, and lubricate the bell crank assembly portion of the bypass control mechanism.

Although the 18 month PM, along with other maintenance, was performed on the TSV-2 bypass control mechanism in the May 2007 refueling outage, this did not include disassembly and inspection of the bell crank mechanism. The failure of Entergy to have an adequate PM strategy for TSV-2 constituted a performance deficiency.

Analysis:

This finding is more than minor because it is associated with the Equipment Performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, during troubleshooting of the degraded condition of the TSV-2 bypass control mechanism, an automatic reactor scram occurred. The inspectors conducted a Phase 1 screening of the finding in accordance with IMC 0609, Appendix A. The finding was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor scram and the likelihood that mitigating equipment or functions would not be available.

The finding has a cross-cutting aspect in the area of Human Performance, Resources component, because Entergy did not maintain an effective PM program for TSV-2.

Specifically, the PM for TSV-2 did not specify a periodic activity to inspect, rebuild, and lubricate the bell crank assembly portion of the bypass control mechanism. H.2(c)

Enforcement:

No violation of regulatory requirements occurred, because the event involved non-safety related systems. Entergy entered this condition into their corrective action program (CR-2007-3349). Corrective actions identified in the root cause report include creation of a PM activity to rebuild the bypass control mechanism; inspection, rebuild, and lubrication of the TSV-2 bell crank and bypass control assemblies during the next refueling outage; and a review of turbine valve PMs to ensure linkages and bell cranks are properly lubricated and rebuilt at intervals. Since this issue is of very low safety significance and has been entered into Entergys corrective action program, it is classified as a finding. FIN 05000271/2007004-02, Reactor Scram During Troubleshooting Due to Inadequate Main TSV Preventive Maintenance

1R13 Maintenance Risk Assessment and Emergent Work Evaluation

a. Inspection Scope

(6 samples)

The inspectors evaluated on-line risk management for four planned maintenance activities and two emergent repair activities. The inspectors reviewed maintenance risk evaluations, work schedules, corrective actions, and control room logs to determine if concurrent or emergent maintenance activities significantly increased the plant risk. The inspectors compared reviewed items and activities to requirements listed in AP 0125, "Plant Equipment," and AP 0172, "Work Schedule Risk Management - Online." A list of documents reviewed is provided in the Attachment of this report. The inspectors walked down areas of the plant containing equipment that was determined to have higher risk significance during the following work activities:

• Planned maintenance on the B RHR system and orange risk configuration;

• Planned maintenance on the C RHR pump discharge check valve and yellow risk configuration;

• Planned ACS deep basin draining and inspection and yellow risk configuration;

• Planned maintenance for the week of August 20 and green risk configuration;

• Emergent work associated with the C SW pump and yellow risk configuration; and

• Emergent work associated with the A SW pump and yellow risk configuration.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

(5 samples)

The inspectors reviewed five operability evaluations associated with degraded or non-conforming conditions to ensure that operability and functionality was justified. The inspectors evaluated the operability evaluations against the guidance contained in NRC Regulatory Issue Summary 2005-20, Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability, as well as Entergy procedure ENN-OP-104, Operability Determinations. A listing of documents reviewed is provided in the Attachment of this report. The inspectors also discussed the conditions with operators and system and design engineers, as necessary. The following evaluations were reviewed.

• Leak in C SW discharge piping;

• Fire header pressure lowered unexpectedly;

• Condensate storage tank (CST) temperature high out of band;

• CST leakage; and

• Cooling tower cell 2-1 operability

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications

a. Inspection Scope

(9 samples)

The inspectors reviewed nine permanent plant modification packages to determine if the design bases, licensing bases, and performance capability of risk significant SSCs had been degraded through plant modifications. The inspectors performed walkdowns of selected plant systems and components, interviewed plant staff, and reviewed applicable documents, including procedures, calculations, modification packages, engineering evaluations, drawings, corrective action program documents, the UFSAR, and TS.

The inspectors reviewed selected attributes (component safety classification, energy requirements supplied by supporting systems, seismic qualification, instrument setpoints, uncertainty calculations, electrical coordination, electrical loads analysis, and equipment environmental qualification) to determine if they were consistent with the design and licensing bases. Design assumptions were reviewed to evaluate if they were technically appropriate and consistent with the UFSAR. For each modification, the 10 CFR 50.59 screenings or evaluations were reviewed, as described in section 1R02 of this report. The inspectors reviewed procedures, calculations, and the UFSAR to determine if they were properly updated with revised design information and operating guidance. The inspectors also reviewed the as-built configuration to determine if it was accurately reflected in the design documentation and that post-modification testing was adequate to ensure the SSCs would function properly. The inspectors also reviewed issues that had been entered into the corrective action program to determine whether Entergy had been effective in identifying and resolving problems associated with the plant modification process and activities. A listing of modifications and documents reviewed is provided in the Attachment to this report.

b. Findings

No findings of significance were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

(5 samples)

The inspectors reviewed post-maintenance testing (PMT) activities on risk-significant systems. The inspectors either observed the PMT or reviewed completed PMT documentation to evaluate if the test data met the required acceptance criteria contained in the applicable WO, TS, UFSAR, and/or inservice testing program. Where testing was directly observed, the inspectors verified that installed test equipment was appropriate and controlled and that the test was performed in accordance with applicable station procedures. The inspectors also verified that the test activities were adequate to ensure system operability and functional capability following maintenance, systems were properly restored following testing, and any discrepancies were appropriately documented in the corrective action program. The inspectors reviewed the PMTs performed for the following maintenance activities:

• Testing of the B RHR pump suction from recirculation loop A valve (V10-15B),

PMT in accordance with WO 51191102;

• Mechanical troubleshooting #2 main TSV, PMT in accordance with WO 121155;

• Inspection and repair of C RHR pump discharge check valve (V10-48C), PMT in accordance with WO 51072384;

• Repair of C SW discharge pressure indicator piping leak, PMT in accordance with WO 115614; and

• Replacement of control rod hydraulic control unit (HCU) 06-31 accumulator, PMT in accordance with WO 51079490.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

(1 sample)

The inspectors evaluated the below forced outage activities, after the August 30, 2007, scram, to determine if Entergy considered risk when developing outage schedules; adhered to administrative risk reduction methodologies for plant configuration control; and adhered to their operating license, TS requirements, and approved procedures:

• Monitoring of Shutdown Activities - The inspectors observed the shutdown of the reactor plant. The plant remained in hot shutdown throughout the forced outage;

• Control of Outage Activities - The inspectors reviewed the daily shutdown risk assessment and work schedule to evaluate if Entergy appropriately addressed risk; and

• Startup Activities - The inspectors observed portions of the startup of the reactor plant, including criticality and power increase, following the completion of the forced outage.

The inspectors also evaluated whether Entergy identified problems related to the forced outage and entered them into their corrective action program. Specifically, the inspectors evaluated if Entergy identified the cause of the scram and took appropriate corrective actions prior to restarting the reactor. A listing of documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

(5 samples)

The inspectors observed surveillance testing to determine if the test acceptance criteria specified for each test was consistent with TS and UFSAR requirements, the test was performed in accordance with the written procedure, the test data was complete and met procedural requirements, and the system was properly returned to service following testing. The inspectors observed selected pre-job briefs for the test activities. The inspectors also determined if discrepancies were appropriately documented in the corrective action program. A listing of documents reviewed is provided in the Attachment of this report. The inspectors verified that the following surveillance testing activities met the above requirements:

• B emergency diesel generator (EDG) monthly slow start operability test in accordance with OP 4126, Section B;

• RCIC comprehensive testing in accordance with OP 4121, Section C;

• A EDG six month fast start operability test in accordance with OP 4126, Section F;

• A core spray (CS) pump quarterly operability test in accordance with OP 4123, Section C; and

• A CS quarterly motor operated valve/injection check valve closure test in accordance with OP 4123, Section B.

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications

a. Inspection Scope

(1 sample)

The inspectors reviewed temporary modification (TM) EC-2428, Temporary Modification to Eliminate Ground On DC-1." The inspectors reviewed the TM package to determine if it adversely affected the availability, reliability, or functional capability of any risk-significant SSCs. The inspectors reviewed the installation and control of the TM against the TS, UFSAR, and Entergy Corporate Procedure ENN-DC-136, Temporary Alterations, to evaluate if the license and design basis were properly maintained. A listing of documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas (71121.01)

a. Inspection Scope

(13 samples)

During September 10-13, 2007, the inspector conducted the following activities to determine if Entergy was properly implementing physical, engineering, and administrative controls for access to high radiation areas and other radiologically controlled areas, and that workers were adhering to these controls when working in these areas. Implementation of the access control program was reviewed against the criteria contained in 10 CFR 20, TS, and Entergy procedures. A listing of documents reviewed is provided in the Attachment of this report.

The inspector walked down exposure significant work areas of the plant (reactor building, turbine building, radwaste building, and auxiliary off-gas building) and reviewed Entergy controls and surveys to determine if Entergy surveys, postings, and barricades were acceptable and in accordance with regulatory requirements.

The inspector walked down exposure significant work areas of the plant and, with the assistance of a radiation protection technician, conducted independent surveys to determine whether prescribed radiation work permit and procedural controls were in place and whether Entergy surveys and postings were complete and accurate.

The inspector reviewed Entergys physical and programmatic controls for highly activated materials stored underwater in the spent fuel pool and evaluated through observation and a review of the applicable access control procedure.

The inspector reviewed Entergy radiation protection program self-assessments and audits during 2007 to determine if identified problems were entered into the corrective action program for resolution.

Ten CRs associated with the radiation protection access control and as low as reasonably achievable (ALARA) areas between May 2007 and August 2007, were reviewed and discussed with Entergy staff to determine if the follow-up activities were being conducted in an effective and timely manner commensurate with their safety significance. Based on the CRs reviewed, repetitive deficiencies were screened to determine if the Entergys self-assessment activities were identifying and addressing these deficiencies.

There were no Occupational Exposure Performance Indicator incidents during the current assessment period.

Changes to the high radiation area and very high radiation area procedures since the last inspection in this area were reviewed and management of these changes was discussed with the Radiation Protection Manager.

Controls associated with potential changing plant conditions, to anticipate timely posting and controls of radiation hazards, were discussed with a radiation protection supervisor.

All accessible locked high radiation area entrances in the plant were examined to determine if they were locked, including the key inventory. Also, the inspector reviewed the controls for the locked and very high radiation area keys.

Several CRs were reviewed to evaluate if the incidents were caused by radiation worker errors and to determine if there were any trends or patterns and if the Entergys corrective actions were adequately addressing these trends.

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

(2 samples)

During September 10-13, 2007, the inspector conducted the following activities to determine if Entergy was properly maintaining individual and collective radiation exposures ALARA. Implementation of the ALARA program was reviewed against the criteria contained in 10 CFR 20.1101(b) and Entergy procedures. A listing of documents reviewed is provided in the Attachment of this report.

The process for adjusting work activity exposure estimates was evaluated for emergent work and unexpected radiological conditions. The methodology for the exposure estimate adjustments was evaluated with respect to radiation protection and ALARA principles.

There was one declared pregnant worker during the current assessment period. The applicable personnel exposure results and monitoring controls employed by Entergy were reviewed with respect to the requirements of 10 CFR 20.

b. Findings

No findings of significance were identified.

Cornerstone: Public Radiation Safety (PS)

2PS3 Radiological Environmental Monitoring Program (REMP) (71122.03)

a. Inspection Scope

(10 samples)

The inspector reviewed the current Annual Radiological Environmental Operating Report, and Entergy assessment results, to determine if the REMP was implemented as required by TS and the offsite dose calculation manual (ODCM). The review included changes to the ODCM with respect to environmental monitoring commitments in terms of sampling locations, monitoring and measurement frequencies, land use census, interlaboratory comparison program, and analysis of data. The inspector reviewed the ODCM to identify environmental monitoring stations. In addition, the inspector reviewed the following: Entergy self-assessments and audits, event reports, interlaboratory comparison program results, the UFSAR for information regarding the environmental monitoring program and meteorological monitoring instrumentation, and the scope of the audit program to determine if it met the requirements of 10 CFR 20.1101.

The inspector walked down seven air particulate and iodine sampling stations; one south river water sampling station; three dairy farms; and 25 thermoluminescent dosimeter (TLD) monitoring locations to determine if they were located as described in the ODCM, and to evaluate if the equipment material condition was acceptable.

The inspector observed the collection and preparation of a variety of environmental samples to determine if the environmental sampling was representative of the release pathways as specified in the ODCM, and if sampling techniques were in accordance with procedures.

Based on direct observation and review of records, the inspector evaluated whether the primary and backup meteorological tower instruments were operable, calibrated, and maintained in accordance with the guidance contained in the UFSAR, NRC Safety Guide 23, and Entergy procedures. The inspector reviewed the meteorological data readout and recording instruments in the control room and at the tower to determine if they were operable.

The inspector reviewed each event documented in the Annual REMP which involved a missed sample, inoperable sampler, lost TLD, or anomalous measurement for the cause and corrective actions. The inspector conducted a review of Entergys assessment of any positive sample results.

The inspector reviewed any significant changes made by Entergy to the ODCM as the result of changes to the land census or sampler station modifications since the last inspection. The inspector also reviewed technical justifications for any changed sampling locations to determine if Entergy performed the required reviews to ensure that the changes did not affect its ability to monitor the impacts of radioactive effluent releases on the environment.

The inspector reviewed the calibration and maintenance records for air samplers. The inspector reviewed the following: the results of Entergys interlaboratory comparison program to evaluate the adequacy of environmental sample analyses performed by Entergy, the quality control evaluation of the interlaboratory comparison program and the corrective actions for any deficiencies, the determination of any bias to the data and the overall effect on the REMP, and quality assurance audit results of the program to determine whether Entergy met the TS/ODCM requirements. The inspector evaluated the detector sensitivities for counting samples to determine if they in accordance with the TS/ODCM and reviewed the results of the quality control program, including the interlaboratory comparison program, to evaluate the adequacy of the program.

The inspector observed the radioactive material survey and release locations and inspected the methods used for control, survey, and release. The inspector also observed the performance of personnel surveying and releasing material for unrestricted use to determine if the work was performed in accordance with plant procedures.

The inspector reviewed the radiation monitoring instrumentation used for the release of material from the radiological controlled area to determine if it was appropriate for the radiation types present and was calibrated with appropriate radiation sources. The inspector reviewed Entergys equipment to ensure the radiation detection sensitivities were consistent with the NRC guidance contained in Circular 81-07 and Information Notice 85-92 for surface contamination and HPPOS-221 for volumetrically contaminated material.

The inspector reviewed Entergys audits and self-assessments related to the radiological environmental monitoring program since the last inspection to determine if identified problems were entered into the corrective action program, as appropriate. Selected CRs were reviewed since the last inspection to determine if identified problems accurately characterized the causes and corrective actions were assigned to each commensurate with their safety significance. Any repetitive deficiencies were assessed to determine if Entergys self-assessment activities were identifying and addressing these deficiencies.

A listing of documents reviewed is provided in the Supplemental Information attachment of this report.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

a. Inspection Scope

(3 samples)

Mitigating Systems Cornerstone The inspectors reviewed Entergy submittals for the Safety System Functional Failures performance indicator (PI). The PI definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, and AP 0094, NRC Performance Indicator Reporting, were used to verify the completeness of the PI data reported. The inspectors reviewed the licensee event reports (LERs), operator logs, and maintenance rule out of service logs to verify the accuracy and completeness of the PI data for the period from October 1, 2006, through June 30, 2007.

Occupational Radiation Safety Cornerstone The inspector reviewed implementation of the Entergys Occupational Exposure Control Effectiveness PI Program. Specifically, the inspector reviewed CRs and radiologically controlled area dosimeter exit logs for the past four calendar quarters. These records were reviewed for occurrences involving locked high radiation areas, very high radiation areas, and unplanned exposures against the criteria specified in NEI 99-02 to verify that all occurrences that met the NEI criteria were identified and reported as performance indicators.

Public Radiation Safety Cornerstone The inspector reviewed a listing of relevant effluent release reports for the past four calendar quarters, for issues related to the RETS/ODCM Radiological Effluent PI against the criteria specified in NEI 99-02 . The inspector reviewed the monthly projected dose assessment results due to radioactive liquid and gaseous effluent releases, quarterly projected dose assessment results due to radioactive liquid and gaseous effluent releases, and dose assessment procedures to verify the accuracy and completeness of the PI data.

b. Findings

No findings of significance were identified.

4OA2 Problem Identification and Resolution

.1 Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems, the inspectors performed a screening of each item entered into Entergys corrective action program. This review was accomplished by reviewing printouts of each CR, attending daily screening meetings and/or accessing Entergys database. The purpose of this review was to identify conditions such as repetitive equipment failures or human performance issues that might warrant additional follow-up.

b. Findings

No findings of significance were identified.

.2 Annual Sample: Dropped Fuel Pin (1 Sample)

a. Inspection Scope

The inspectors reviewed Entergys actions in response to the dropped fuel pin during gamma scanning on May 26, 2007. The Gamma Scan was being done on irradiated fuel rods stored in the spent fuel pool. The inspectors reviewed the CR, including the RCA, the extent of condition, the corrective actions, and the applicable operating procedures. The inspectors also interviewed the responsible reactor engineer.

b. Findings

No findings of significance were identified. The Entergy investigation, and the associated RCA report, appeared thorough. The description of the event included sufficient detail to support Entergys conclusions. The gamma scan equipment performed as designed to capture the dropped fuel pin, subsequent investigation of the fuel pin showed no signs of damage. All personnel associated with the event were qualified fuel handlers. The root cause of the dropped fuel pin was attributed to an inadequate process to compensate for a clamp design weakness; the clamp allowed two possible means to grab the fuel rod. The fuel handlers were not aware of this vulnerability.

4OA3 Event Followup

.1 Operator Response following a C Feedpump Trip

a. Inspection Scope

(1 sample)

On August 4, 2007, with the plant at 100 percent power, operators responded to an unexpected trip of the C reactor feed pump. During the transient, an automatic reactor recirculation runback occurred reducing power to approximately 80 percent, as designed. The operators took prompt action to insert control rods, per a predetermined rapid shutdown sequence, to reduce reactor power to within the maximum extended load line limit analysis power to flow boundary, in accordance with station procedures.

The inspectors evaluated the adequacy of operator actions in response to the reactor feed pump trip. A listing of documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance were identified.

.2 Loss of Control Board Annunciators

a. Inspection Scope

(1 sample)

On July 31, 2007, operators identified that a portion of control room alarm panel 9-3 would not light when tested and the associated alarm horn for the panel did not sound.

The affected alarms were related to radiation process and area conditions. The resident inspectors responded to the control room to determine if operations had properly considered each alarm function lost in reference to the individual alarm response procedures. The inspectors reviewed the determination that no TS action statement, Emergency Action Level, or NRC reportability criteria had been met. Operations established compensatory actions involving hourly local monitoring of the affected radiation parameters. The inspectors reviewed the compensatory actions to evaluate if design assumptions remained valid. On August 2, 2007, after a power supply was replaced, the annunciators were returned to service and compensatory measures discontinued. A listing of documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance were identified.

.3 Partial Collapse of a Non-Safety-Related Cell in the West Cooling Tower

a. Inspection Scope

(1 sample)

On August 21, 2007, at 3:30 p.m., with the plant at 100 percent power, a non-safety related portion (cell 2-4) of the west CT collapsed. The resident inspectors responded to the control room and assessed the operators response to the event, and locally observed the CT structural damage. Immediately following the event, the plant continued to operate at 100 percent power and all associated safety-related SSCs remained operable. At approximately 4:00 p.m., the operators isolated circulating water flow to the CT and secured the cooling fans, the operators then reduced power to approximately 35 percent to comply with State thermal discharge limits. The inspectors observed these activities to evaluate Entergys response to determine if it was consistent with station procedures and training. A listing of documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance related to event response were identified. A performance deficiency associated with the cause of the unplanned power reduction is documented in Section 1R12.1.

.4 Operator Response to an Automatic Reactor Scram

a. Inspection Scope

(1 sample)

On August 30, at 3:13 p.m., with the Unit at 63 percent power, all four main turbine stop valves unexpectedly closed during troubleshooting activities on TSV-2. As a result, an automatic reactor scram occurred. The resident inspectors responded to the control room to determine if operator response was in accordance with station procedures and training and that mitigating systems responded as expected. The inspectors evaluated whether Entergy appropriately considered emergency action level entry and notification requirements. In addition, the inspectors reviewed operator logs, plant computer data, CRs, and the Post Trip Review report. A listing of additional documents reviewed is provided in the Attachment of this report.

b. Findings

No findings of significance related to event response were identified. A performance deficiency associated with the cause of the scram is documented in Section 1R12.2.

Failed to Open (1 sample)

On June 8, 2007, with the reactor at 81 percent power, Entergy identified that the HPCI pump injection valve (V23-19) did not open on a manual signal from the control room during a surveillance test. Entergy entered the condition into their corrective action program and a root cause evaluation was performed. Entergy determined that one of the motor operated valve (MOV) contacts (72/C) was in the intermediate position, causing electrical and mechanical interlocks that prevented the open contactor (82/O)from energizing. Entergy identified that the 72/C contacts were pitted and worn, causing the contact surfaces to overheat and weld together. Entergy determined that the PM performed on the valve control circuitry was inadequate, in that it did not contain sufficient guidance on how to determine contact wear and when the contacts should be replaced. The inadequate PM activity constituted a performance deficiency.

This finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective of assuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

In accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the Phase 1 screen required a Phase 2 analysis, because the inoperability of the HPCI system represented a loss of system safety function. The inspector conducted a Phase 2 SDP analysis, using the following assumptions, and the Risk-Informed Inspection Notebook for Vermont Yankee Nuclear Power Station, Revision 2: the exposure time was approximately six days and no operator recovery credit provided. Based upon the SDP Notebook Pre-Solved Excel Spreadsheet, this finding was preliminarily assessed as greater than Green.

Subsequently, a Region I Senior Reactor Analyst (SRA) performed a Phase 3 evaluation using the Vermont Yankee Standardized Plant Analysis Risk (SPAR) model using the same assumptions stated above. The SPAR model internal events risk assessment yielded a low E-7 increase in core damage frequency (CDF). Based upon the internal risk contribution being greater than 1E-7, the SRA reviewed the finding for external event risk contributors, consistent with IMC 0609, Appendix A, Attachment 3. No fire or flooding risk contributors were identified and the seismic risk contribution (principally seismically induced loss of offsite power events) was determined to be of minimal consequence (less than 1E-11). Accordingly, the total (internal and external) increase in CDF for this finding was low E-7, or very low safety significance (Green). The dominant postulated core damage scenarios involved a loss of the Division 2 Direct Current Bus followed by the failure to make-up (high pressure or low pressure injection, following depressurization) to the reactor vessel. Neither the RCIC system nor the low pressure injection systems, in conjunction with the automatic depressurization system, were adversely impacted by this finding and were available for accident mitigation while the HPCI system was degraded.

The issue was entered into Entergys corrective action program (CR 2007-2372).

Corrective actions included: performing an extent of condition review to identify affected contacts in other systems, evaluating system operability, developing a prioritized replacement schedule based on risk significance, developing criteria for replacement during PM activities, and requiring periodic replacement of related heavily loaded MOV contactors. The inspectors reviewed the LER, root cause evaluation, extent of condition review, and corrective actions. A listing of documents reviewed is provided in the of this report.

This licensee-identified finding involved a violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The enforcement aspects of the violation are discussed in Section 4OA7. This LER is closed.

4OA5 Other Activities

Strike Contingency Plan (92709)

a. Inspection Scope

Entergy developed a Management Alternative Plan (MAP) to ensure a sufficient number of qualified personnel were available to continue Entergy operations in the event that the International Brotherhood of Electrical Workers (IBEW) union personnel engaged in a job action upon the expiration of their contract on August 19, 2007. Using the guidance contained in NRC Inspection Procedure 92709, Licensee Strike Contingency Plans, the inspectors reviewed Entergys plans to address a potential job action at the site.

The inspection included an evaluation of the MAP content and the actions needed to implement the plan; a review to determine whether the number of qualified personnel needed for the proper operation of the facility would be available; a review to determine if reactor operations and station security would be maintained, as required; and, a review to determine if the plan complied with TS requirements and other NRC requirements. On August 17, 2007, the contract was extended and negotiations continued until August 27, 2007, when IBEW union personnel approved a contract. No job action was taken.

b. Findings

No findings of significance were identified.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On August 2, 2007, the radiation protection inspector presented the inspection results to Mr. Norman Rademacher, Director of Engineering, and other members of the VY staff.

The inspectors confirmed that no proprietary information was provided or examined during the inspection.

On August 10, 2007, the permanent plant modifications inspectors presented the inspection results to Mr. William Maguire, General Manager of Plant Operations, and other members of the VY staff. The inspectors confirmed that no proprietary information was provided or examined during the inspection.

On September 14, 2007, the radiation protection inspector presented the inspection results to Mr. David Mannai, Licensing Manager, and other members of the VY staff.

The inspectors confirmed that no proprietary information was provided or examined during the inspection.

On October 9, 2007, the resident inspectors presented the inspection results to Mr. Theodore Sullivan, Site Vice President, and other members of the VY staff. The inspectors confirmed that no proprietary information was provided or examined during the inspection.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by Entergy and is a violation of NRC requirements which meet the criteria in Section VI.A.1 of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV. A listing of documents reviewed is provided in the Attachment of this report.

10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states that Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.

Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Contrary to the above, Entergy did not have a maintenance procedure describing appropriate acceptance criteria for determining when to replace the HPCI MOV contacts. As a result, HPCI was inoperable for approximately six days after MOV V23-19 failed to open during a planned surveillance. This finding was entered into Entergys corrective action program (CR 2007-2372). This finding is of very low safety significance (Green) as determined by a NRC Phase 3 significance determination (Section 4OA3.5).

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

J. Dreyfuss, Director of Nuclear Safety

W. Maguire, General Manager of Plant Operations

D. Mannai, Licensing Manager

N. Rademacher, Director of Engineering

T. Sullivan, Site Vice President

G. Von Der Esch, Acting Operations Manager

S. Wender, Radiation Protection Manager

LIST OF ITEMS

OPENED AND CLOSED

Opened and Closed

05000271/2007004-01 NCV Inadequate Inspection Program Resulted in a Partial Collapse of a Non-Safety-Related Cooling Tower Cell (Section 1R12.1)

05000271/2007004-02 FIN Reactor Scram During Troubleshooting Due to Inadequate Main TSV Preventive Maintenance (Section 1R12.2)

Closed

05000271/2007002-00 LER High Pressure Coolant Injection (HPCI) System Valve Failed to Open (Sections 4OA3.5 and 4OA7)