IR 05000271/2010003

IR 05000271-10-003, on 04/01/2010 - 06/30/2010; Vermont Yankee Nuclear Power Station; Refueling and Other Outage Activities

ML102100320
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	07/29/2010
From:	Diane Jackson NRC/RGN-I/DRP/PB5
To:	Michael Colomb Entergy Nuclear Operations
References
IR-10-003
Download: ML102100320 (42)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION I

475 ALLENDALE ROAD KING OF PRUSSIA, PA 19406-1415 July 29, 2010 Mr. Michael Colomb Site Vice President Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station Vernon, VT 05354 SUBJECT: VERMONT YANKEE NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000271/2010003

Dear Mr. Colomb:

On June 30, 2010, the U.S. NUclear Regulatory Commission (NRC) completed an inspection at your Vermont Yankee Nuclear Power Station. The enclosed inspection report documents the inspection results, which were discussed on July 26, 2010, with you and other members of your staff.

The inspection examined activities performed under your license as they relate to safety and compliance with the Commission's rules and regulations, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents one self-revealing finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requirements. However. because of the very low safety significance and because it has been entered into your corrective action program (CAP), the NRC is treating this finding as a non-cited violation (NCV). consistent with Section VLA.1 of the NRC's Enforcement Policy. If you contest any NCV, 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 Senior Resident Inspector at Vermont Yankee. In addition, if you disagree with the cross cutting aspect assigned to the 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 I, and the NRC Senior Resident Inspector at Vermont Yankee. 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}.

n'(~~-

Donald E. J k n, Chief Reactor Pro s Branch 5 Division of Reactor Projects Docket No. 50-271 License Nos. DPR-28

Enclosure:

Inspection Report No. 05000271/2010003

. wI Attachment: Supplemental Information

REGION I==

Docket No.: 50-271 license No.: DPR-28 Report No.: 05000271/2010003 licensee: Entergy Nuclear Operations. Inc.

Facility: Vermont Yankee Nuclear Power Station i .

Location: Vernon, Vermont 05354-9766 Dates: April 1, 2010 through June 30, 2010 Inspectors: D. Spindler, Sr. Resident Inspector, DRP H. Jones, Resident Inspector, ORP J. Noggle, Sr. Health Physicist, DRS T. Burns, Reactor Inspector, DRS M. Reichard, Health Physicist. DNMS Approved by: Donald E. Jackson, Chief Projects Branch 5 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000271/2010003; 04/01/2010 - 06/30/2010; Vermont Yankee Nuclear Power Station;

Refueling and Other Outage Activities.

This report covered a three-month period of inspection by resident inspector staff and region based inspectors. One Green, self-revealing finding, which was determined to be a non-cited violation (NCV), was identified. The significance of most findings is indicated by their co!or . I (Green, White, Yellow, Red) using Inspection Manua! Chapter (IMC) 0609, "Significance Determination Process." The cross-cutting aspect for the finding was determined using IMC 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.

Cornerstone: Mitigating Systems

Green.

A self-revealing, NCV of very low safety significance (Green) of Technical Specification (TS) 6.4, "Procedures," was identified when operators inadvertently drained water from the reactor pressure vessel (RPV) during integrated emergency core cooling system (ECGS) testing. Specifically, Entergy failed to establish the initial plant conditions necessary to perform integrated ECCS testing without causing an inadvertent drain down of the vessel through the main steam lines, the RCIC turbine, and into the torus. Entergy restored the RPV inventory, initiated a CR to perform a root cause evaluation of the issue, and assigned a corrective action to revise the procedure in order to preclude recurrence in future outages.

The inspectors determined that the finding was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events Cornerstone, and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the implementation of the inadequate procedure guidance resulted in an unexpected loss of RPV water inventory of approximately 2100 gallons.

The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Resources component, because the test procedure was inadequate. Specifically, the procedure did not provide adequate directions for establishing plant conditions during a test that had the capability of draining RCS inventory H.2(c). (Section 1R20)

REPORT DETAILS

Summary of Plant Status

Vermont Yankee (VY) Nuclear Power Station began the inspection period operating at 100 percent power. On April 24, 2010, VY performed a planned power reduction to support the refueling and maintenance outage. On May 26, 2010, VY began power ascension after completion of the refueling outage; however, the unit automatically shut down at 70 percent power due to a recent modification design issue. After correcting the modification issue, VY began power ascension on May 29, 2010. VY reached 100 percent power on June 1, 2010, and remained at or near 100 percent power for the duration of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

.1 Grid Stability - Summer Readiness of Offsite and Alternate AC Power Systems

a.

Inspection Scoge The inspectors performed a review of Entergy's offsite and alternate AC power system readiness for adverse weather. The inspectors reviewed Entergy's plant features and procedures for operation and continued availability of their AC power systems to determine if they were appropriate. The inspection focused on Entergy's procedures for communication protocols with the transmission system operator (TSO) to determine if appropriate information would be exchanged when issues arise that could impact the offsite power system. The inspectors also reviewed Entergy's procedures to determine if they addressed necessary actions to be taken if notified by the TSO that they needed to transfer safety-related loads to the onsite power supply, compensatory actions to be taken if it were not possible to predict grid conditions, and required communications between Entergy and the TSO. The inspectors interviewed personnel and performed a walkdown of the switchyard. A list of documents reviewed is provided in the

.

b. Findings

No findings were identified.

1R04 Equipment Alignment

Partial Equipment Alignment a.

Inspection ScoQe The inspectors performed four partial system walk downs to verify correct 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 piping and instrumentation drawings, and operating procedures (OPs). The inspectors verified valve positions and the general condition of selected components. Finally. the inspectors evaluated material condition, housekeeping, and component labeling. The documents reviewed are listed in the Attachment. The following systems were inspected:

• 'A' train residual heal removal service water (RHRSW) system during 'B' train RHRSW system testing;

• 'A' train residual heat removal (RHR) during 'B' train RHR testing;

• 'S' train emergency diesel generator (EDG) system during 'A' train EDG system maintenance; and

• Advanced off-gas system (AOG) after maintenance.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Quarterly Inspection

a.

Inspection ScoQe The inspectors performed inspections of six fire areas based on a review of the Vermont Yankee Safe Shutdown Capability Analysis, the Fire Hazards Analysis, and the Individual Plant Examination for External Events (IPEEE). The inspectors reviewed Entergy's fire protection program to determine the specified fire protection design features, fire area boundaries, and combustible loading requirements for the selected areas. The inspectors verified, consistent with applicable administrative procedures, that combustibles and ignition sources were adequately controlled; passive fire barriers, manual fire-fighting equipment, and detection and suppression equipment were appropriately maintained; and compensatory measures for out-ot-service, degraded, or inoperable fire protection equipment were implemented in accordance with Entergy's fire protection program. The inspectors evaluated the fire protection program for conformance with the requirements of License Condition 3.F. The documents reviewed are listed in the Attachment. The following fire areas were inspected:

• Fire area FA-8, elevation 252', 'A' EDG;

• Fire area FA-9, elevation 252', 'B' EDG;

• Fire zone FZ-7. elevation 228', condenser bay basement;

• Fire zone FZ-?, elevation 248', condenser bay and ground floor;

• Reactor building RB-4, elevation 252', reactor building south: and

• Reactor building RB-4, elevation 213', high pressure coolant injection (HPCI).

b. Findings

No findings were identified .

.2 Annual Inspection

The inspectors evaluated Entergy's fire brigade performance by observing an announced fire brigade drill on April 2. 2010. The inspectors verified that the number of individuals assigned to the fire brigade response met the minimum specified number; each member donned the protective clothing, tumout gear and self-contained breathing apparatus; the fire brigade leader exhibited command of the fire brigade and had copies of the pre-fire plans; the fire brigade arrived at the fire scene in a timely manner; the licensee's drill scenario was followed and acceptance criteria for the drill objectives were met; and the licensee performed a post-drill critique to discuss any failures and weaknesses associated with the fire drill performance.

b. Findings

No findings were identified.

1RO? Heat Sink Performance (71111.07 - 2 samples) Annual Inspection

a. Inspection Scope

The inspectors reviewed the results of the thermal performance tests of emergency diesel generator heat exchangers for DG-1-1A and DG-1-1B including the scavenging air cooler. lubricating oil cooler, and jacket water cooler for each diesel. The inspectors discussed the test results with the system engineer and reviewed the completed surveillance data to determine whether test results met acceptance criteria, which considered differences between test and design basis accident conditions. The inspectors also reviewed Entergy's corrective action program (CAP) to ensure significant heat exchanger performance problems were appropriately identified and documented, and that corrective actions assigned, if any, were appropriate. A list of documents reviewed is provided in the Attachment to this report.

b. Findings

No findings were identified.

1R08 Inservice Inspection (lSI)

a. Inspection Scope

The purpose of this inspection was to assess the effectiveness of the licensee's lSI program for monitoring degradation of reactor pressure vessel internals, reactor coolant system boundary, risk significant piping system boundaries, and the containment boundary. The inspector assessed the inservice inspection activities using requirements and acceptance criteria for component examination specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, and applicable NRC Regulatory Requirements.

The inspector selected a sample of nondestructive examination (NDE) activities to perform a documentation review of those NDE activities for compliance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI. The sample selection was based on the inspection procedure objectives, sample availability, and risk priority of those components and systems where degradation could result in a significant increase in risk of core damage. The inspector verified by documentation review that test procedures and examiner qualifications had been reviewed and approved for use by the licensee. The inspector reviewed a sample of these procedures and qualifications to determine they were current and in accordance with the ASME Code requirements. In addition, the inspector performed this review to determine that examiners had been trained and qualified for use of the performance demonstration initiative (PDI) manual ultrasonic test procedures. The inspector selected a sample of Indication Notification Reports (INR) and Condition Reports (CRs) to evaluate the licensee's effectiveness in the identification and resolution of relevant indications discovered during the observed lSI activities. The inspector's documentation review of non-destructive testing included the following:

• Ultrasonic test (UT), manual PDI-UT of weld overlay applied to core spray nozzles N5A-SE and N5B-SE (Examination data was reviewed for comparative evaluation with the UT examinations performed in 1993, 1998 and 2004);

• Magnetic particle test (MT) of sweep weld RH17-S-349A, to the RHR piping, drawing G191172, lSI Isometric ISI-:5920-9212;

• Visual examination (VT-3) of component supports on torus suppression chamber (NB-H1951-TSS10, drawings 5920-9144,6135,9131 and 9133); and

• Liquid Penetrant test (PT) of nozzle to safe end dissimilar butt weld N10-SE, drawing ISI-RPV-103, location elevation 269 ft, azimuth 350 degrees.

The inspector reviewed the visual inspection results of various in-vessel components including jet pump structural members, leveling devices, miscellaneous fasteners and tack welds made to secure components in place. These observations were made to determine the examiner skill, and to assess the test equipment performance, examination technique, and the quality of the inspection environment (water clarity).

The inspector performed a review of in-vessel component non-conforming conditions identified in CR's 2010-02011,2010-02016 and INR-VYR 28-05,28-06,28-07,28-08, 28-09, 28-10 and 28-11. Additional CR's listed in the Attachment to this report were reviewed by the inspector to evaluate the characterization and disposition of relevant indications identified during the in-vessel visual inspection (IWI) of structural components within the vessel.

The inspector selected three ASME Section XI repair/replacement plans-for review where welding on a safety related pressure boundary was performed. The review was performed to evaluate qualification and control of the welding process specified in the work order. Also, to determine that the weld procedures and welders assigned to perform this special process are qualified in accordance with the requirements of ASME Section XI and, that specified weld examinations are done in accordance with the ASME code requirements. The three ASME Section XI repair/replacement work orders reviewed were:

• W0184964-06 was developed for welding of fitting to service water (SW) piping to facilitate installation of valve V70-164C. This welding was of stainless steel base materials welded in line #3/4SW-18A by groove welding using weld procedure specification (WPS) SS-8/8-8, Rev 21. Welds # 48 and 49 added to the service water system as shown on Weld Map 184964-10 Rev O. Applicable Code for acceptance was ASME Section XI, lSI Class 3;

• W0178169-01 was developed to replace fourteen inch "0" service water pump discharge valve V70-2D. Welding was of carbon steel to carbon steel, using WPS CS-1/1-A Rev 3 by groove welding of pipe to valve and pipe to pipe (3 welds) as shown on weld map 178169-1 Rev O. Applicable Code for acceptance was ASME Section XI, lSI Class 3; and

• W0184964-06 was developed to replace pipe to socket weld #47 in service water system, line 3/4 SW-18A. using WPS BM-8/1-A Rev 2. stainless to carbon steel as shown on weld map 184964-10/RO. Applicable Code for acceptance was ASME Section XI, lSI Class 3.

Additionally, the inspector performed a visual evaluation of the primary containment liner and additional structural members attached to the liner to assess the condition of the protective coating. The evaluation included accessible locations on elevations 252' thru 266'. The inspector performed this visual evaluation to determine the extent of any peeling, blistering, coating loss or other damage as a result of corrosion, foreign material impact or lack of maintenance. Condition Report CR-VTY-2010-02337 was initiated to describe the results of this visual examination.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification Program (71111.110 - 1 sample)

Ouarterly Inspection a.

Insgection Scoge The inspectors observed a simulator-based licensed operator requalification (LOR)exam on June 10, 2010. The inspectors assessed the performance of risk significant operator actions, including the use of emergency operating procedures. 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. The inspectors also compared the simulator configuration with the actual control board configuration. Finally, the inspectors verified that evaluators were identifying and documenting crew performance problems. The documents reviewed are listed in the

.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness (71111.120 - 3 samples)

a. Inspection Scope

The inspectors reviewed performance-based problems involving selected in-scope structures, systems and components (SSCs) to assess the effectiveness of the maintenance program. The reviews focused on the following aspects when applicable:

• Proper maintenance rule scoping in accordance with 10 CFR 50.65;

• Characterization of reliability issues;

• Changing system and component unavailability;

• 10 CFR 50.65 (a){1) and (a)(2) classifications;

• Identifying and addressing common cause failures;

• Trending of system flow and temperature values;

• Appropriateness of performance criteria for SSCs classified (a)(2); and

• Adequacy of goals and corrective actions for SSCs classified (a)(1).

The inspectors reviewed the applicable system health reports, maintenance backlogs, and Maintenance Rule basis documents. The documents reviewed are listed in the

. The following structures, systems and components were inspected:

• Residual Heat Removal (RHR) relief valves;

• Core Spray {CS}; and

• 'A' recirculation pump motor generator set scoop tube positioner.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors evaluated seven maintenance risk assessments for planned and emergent maintenance activities to verify that the appropriate risk assessments were performed prior to removing equipment for work. The inspectors reviewed maintenance risk evaluations. maintenance plans, work schedules, and control room logs to determine if concurrent or emergent maintenance or surveillance activities significantly increased the plant risk. The inspectors reviewed risk assessments to determine if they were performed as required by 10 CFR 50.65{a)(4) and implemented in accordance with Entergy's administrative procedures (AP) 0125, "Plant Equipment," and AP 0172, "Work Schedule Risk Management - Online." When emergent work was performed, the inspectors observed activities to determine if plant risk was promptly reassessed and managed. The documents reviewed are listed in the Attachment. The following maintenance activities were inspected:

• On April 28, 2010, increased risk due to removing the 'A' station battery from service for planned maintenance;

• The week of May 3,2010, increased (yellow) risk due to refueling outage activities;

• The week of May 10, 2010, increased risk due to electrical bus outages to support planned maintenance;

• The week of May 17, 2010, increased risk due to vessel drain down activities, installation of the reactor vessel head, installation of the drywell vessel head, and integrated leak rate testing to support outage activities;

• On June 8, 2010, increased (yellow) risk due to primary containment being declared inoperable;

• On June 10, 2010, high risk evolution associated with the CS system; and

• On June 24, 2010, high risk evolution associated with the 'A' recirculation pump motor~generator set scoop tube repair activities.

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed five operability evaluations associated with degraded or non conforming conditions to assess the acceptability of the evaluations, the use and control of applicable compensatory measures, and compliance with Technical Specifications.

The inspectors reviewed and compared the technical adequacy of the evaluations with the Technical Specifications, Updated Final Safety Analysis Report, associated design basis documents, and Entergy's procedure EN-OP-104, "Operability Determinations."

The documents reviewed are listed in the Attachment. The inspectors reviewed evaluations of the following degraded or non-conforming conditions:

• CR 2010-01686, pitting found on the RHRSW pump P-8-1C discharge piping;

• CR 2010-01821, drywell dome gasket damage;

• CR 2010-01944, SW strainer pitting below minimum wall thickness;

• CR 2010-02285, 'S' emergency diesel generator cylinder #10 temperature differs from other cylinders by more than 300 degrees Fahrenheit; and

• CR 2010-03202, GNF-2 (nuclear) fuel defect in the spacer springs at certain fuel pin locations.

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Temporary Plant Modifications

a. Inspection Scope

(2 samples)

The inspectors reviewed two temporary modifications to ensure they did not adversely affect the availability, reliability, or functional capability of any risk-significant SSCs and assessed the adequacy of the 10 CFR 50.59 evaluations. The inspectors reviewed the engineering change package, walked down the area, interviewed various personnel, and compared the installation and control of the modification to the procedural requirements.

The inspectors also verified that the installation was consistent with the modification documentation; that the drawings and procedures were updated as applicable; and that the post-installation testing was adequate. The documents reviewed are listed in the

. The following two temporary modifications were reviewed:

• Engineering change (EC) 16583, temporary modification to provide power to 480VAC bus-710ads during a bus-7 outage; and

• EC 22659, temporary modification for the leak repair on the AOG two

(2) inch pipe 2" CNPE-172A.

b. Findings

No findings were identified.

.2 Permanent Plant Modifications

a.

Inspection Scol2e (1 sample)

The inspectors reviewed EC16322, "Replacement of Standby Fuel Pool Cooling Service Water Piping," to ensure it did not adversely affect the availability, reliability, or functional capability of any risk-significant SSCs. The inspectors reviewed the engineering change package and completed post-maintenance testing (PMT) package, and observed the system in operation following the installation of the modification. The documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed ten post-maintenance test (PMT) activities on risk-significant systems. The inspectors reviewed these activities to determine whether test acceptance criteria were clear and consistent with design basis documents. When testing was directly observed, the inspectors determined whether installed test equipment was appropriate and controlled, and whether the test was performed in accordance with 10 CFR Part 50, Appendix B, Criterion XI, ~Test Control," and applicable station procedures. Upon completion, the inspectors performed a walkdown to verify that equipment was returned to the proper alignment necessary to perform its safety function, and evaluated whether conditions adverse to quality were entered into the CAP for resolution. The documents reviewed are listed in the Attachment. The inspectors reviewed the PMTs performed for the following maintenance activities:

• On April 9, 2010, 'B' normal fuel pool cooling pump seal replacement;

• On April 21, 2010, hydraulic control unit HCU-22"35 valve (111 ) replacement;

• On April 22, 2010, 'A' reactor water clean up pump seal replacement;

• On May 3,2010, startup transformer (T-3-1B) refuel outage maintenance activities;

• On May 2, 2010, 'c' service water pump discharge valve replacement;

.* On May 16, 2010, 'A' EDG service water piping replacement;

• On May 18, 2010, post maintenance examination during RPV operational pressure test following RPV relief valve replacement;

• On June 8, 2010, 'B' RHR safety relief valve (SR-10-86B) replacement;

• On June 10, 2010, 'B' core spray discharge piping safety relief valve (SR-14-20B)replacement; and

• On June 22,2010, 'B' EDG fuel rack and governor maintenance.

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors evaluated VY refueling outage (RFO) 28 activities to verify that 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. A list of documents reviewed is provided in the Attachment. The following activities were inspected:

• Review of the Outage Plan and Daily Shutdown Risk Assessments - The inspectors reviewed the RFO 28 shutdown risk assessment to verify that Entergy addressed the outage's impact on defense-in-depth for the five shutdown critical safety functions: electrical power availability, inventory control, decay heat removal, reactivity control, and containment. The daily risk assessments, accounting for schedule changes and unplanned activities, were also periodically reviewed to determine whether adequate defense-in-depth was maintained for each safety function when redundancy was limited, and that planned contingencies were appropriate.

• Monitoring of Shutdown Activities - The inspectors observed the shutdown of the reactor plant including reactor plant cooldown and transition to shutdown cooling operations. As soon as practical following the shutdown, the inspectors performed walkdown inspections of the primary containment.

• Electrical Power - The inspectors reviewed the status and configuration of safety related buses throughout RFO 28. The inspectors performed frequent walkdowns of affected portions of the electrical plant including startup transformers, the auxiliary transformer, and the emergency diesel generators.

• Decay Heat Removal System Monitoring - The inspectors monitored decay heat removal (Le., shutdown cooling) status on a daily basis. Monitoring included daily reviews of residual heat removal system alignment, reviews of spent fuel pool cooling system alignment, and reviews of RCS time-to-boil calculations and results.

• Inventory Control - The inspectors performed daily RCS inventory control reviews including reviews of available injection systems and flow paths to ensure consistency with the outage risk plan.

• Reactivity Control - The inspectors observed reactivity management actions taken by control room operators during refueling eVolutions including procedure place keeping, communications with refueling floor personnel, monitoring of source range nuclear instrumentation, and monitoring of individual control rod positions.

• Containment Closure - The inspectors verified proper primary and secondary containment configuration was maintained throughout the outage. The inspectors performed a primary containment "as-found" inspection and a closeout walkdown prior to final containment closure. Finally, the inspectors performed a walkdown of primary and secondary containment to verify that they had been appropriately reestablished prior to and during startup.

• Refueling Activities - The inspectors observed portions of refueling operations, including fuel handling and accounting in the reactor vessel and spent fuel pool. The inspectors also performed an independent core reload verification of 100 percent of the core.

• Heatup and Startup Activities - The inspectors observed portions of the heatup and startup of the reactor plant following the completion of RFO 28.

• Identification and Resolution of Problems - The inspectors also verified that Entergy identified problems related to refueling activities and entered them into their CAP.

b. Findings

Introduction:

A self-revealing, NCV of very low safety significance (Green) of TS 6.4, "Procedures," was identified when operators inadvertently drained water from the reactor pressure vessel (RPV) during integrated emergency core cooling system (ECCS) testing. Specifically, Entergy failed to establish the initial plant conditions necessary to perform integrated ECCS testing without causing an inadvertent drain down of the vessel through the main steam lines, the RCIC turbine, and into the torus.

Description:

On May 17, 2010, while VY was shutdown for a refueling outage, VY experienced an inadvertent loss of reactor coolant inventory when operators initiated integrated ECCS testing. The vessel head was installed and the vessel was flooded up to the RPV flange.

At the start of the integrated EGCS test, the RCIG and HPCI systems were aligned normally. When the test was initiated, the RCIC and HPCI steam supply isolation valves opened as expected. This provided a path for water to flow from the RPV, through the main steam lines, and into the torus.

The event began at 4:12 a.m. with the reactor water level at 285.07 inches. HPGI immediately isolated on a high steam flow signal caused by the flow of water down the HPCI steam line. The event was terminated when operators closed the RGIG inboard steam isolation valve 55-minutes later at 5:07 a.m. with reactor water level at 261.03 inches. RPV water level lowered approximately 24 inches. The licensee restored the RPV inventory.

This is the first time VY conducted integrated EGCS testing with the reactor pressure vessel head installed. In the past, this testing was done with the head removed and main steam line plugs installed to prevent water flow through the path d,escribed above.

However, past operating experience has shown that in this configuration, the test caused sufficient turbulence inside the reactor vessel to create high airborne contamination levels on the refueling floor. Therefore, VY developed a procedUre to conduct integrated ECCS testing with the RPV head installed.

Analysis:

The inspectors determined that the failure to establish appropriate plant conditions prior to ECCS testing in accordance with TS 6.4 was a performance deficiency that was reasonably within Entergy's ability to foresee and correct, and should have been prevented. Traditional Enforcement did not apply, as the issue did not have actual or potential safety consequence, had no willful aspects, nor did it impact the NRC's ability to perform its regulatory function.

A review of IMC 0612, Appendix E, "Minor Examples," revealed that there were no examples similar to this finding. The inspectors determined that the finding was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events Cornerstone, and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safetY functions during shutdown as well as power operations. Specifically, the implementation of the inadequate procedure guidance resulted in an unexpected loss of RPV water inventory of approximately 2100 gallons. The inspectors performed an initial screening of the finding in accordance with IMC 0609.04, "Phase 1 - Initial Screening and Characterization of Findings. n The inspectors concluded that the finding was a "Primary System Loss of Coolant Accident" initiator contributor that affected the safety of the reactor during the refueling outage. The inspectors then evaluated the significance of the finding using Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process." The inspector determined this issue was of very low safety significance using Appendix G, Attachment 1, "Phase 1 Operational Checklist for Both PWRs and BWRs," and specifically, Checklist 8, "BWR Cold Shutdown or Refueling Operation, Time to Boil> 2 Hours: RCS Level <:: 23' Above Top of Flange."

This determination was based on the fact that the reactor vessel water level would not decrease below the level of the main steam Hnes. The inadvertent draining of the water level to the level of the main steam lines would not significantly impact the shutdown safety functions of decay heat removal and maintaining water level in the reactor core.

The inspectors determined that this finding had a cross-cutting aspect in the HI.Jman Performance cross-cutting area, Resources component, because the test procedure was inadequate. Specifically, the procedure did not provide adequate directions for establishing plant conditions during a test that had the capability of draining ReS inventory H.2(c).

Enforcement:

Vermont Yankee Technical Specification 6.4. states, in part, that written procedures shall be maintained covering activities such as refueling operations, and surveillance and testing requirements. Contrary to the above, on May 17,2010, Entergy performed integrated ECCS testing using a procedure that was not maintained adequately, in that, it did not contain appropriate initial plant conditions prior to testing.

Specifically, the procedure did not verify that a RPV drain down path to the torus through the main steam lines and the RCIC and HPCI turbines did not exist. Entergy restored the RPV inventory, initiated a CR to perform a root cause evaluation of the issue, and assigned a corrective action to revise the procedure in order to preclude recurrence in future outages. This resulted in an RPV coolant loss to the torus equivalent to approximately 24 inches of reactor vessel water level. Because this finding was of very low safety significance and has been entered into the CAP (CR 2010-02757), this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 0500027112010003-01: Loss of RCS Inventory During ECCS Testing Due to Inadequate Procedure.)

1R22 Surveillance Testing

a. Inspection Scope

The inspectors observed ten surveillance tests and/or reviewed test data of selected risk-significant SSCs to determine whether the testing adequately demonstrated equipment operational readiness and the ability to perform the intended safety functions.

The inspectors reviewed selected prerequisites and precautions to determine if they were met, evaluated whether the tests were performed in accordance with the written procedure, determined whether the test data was complete and met procedural requirements, and assessed whether SSCs were properly returned to service following testing. The inspectors also verified that conditions adverse to quality were entered into the CAP for resolution. The documents reviewed are listed in the Attachment. The inspectors reviewed the following surveillance tests:

• On April 20, 2010, '0' RHRSW pump operability and full flow test (1ST);

• On April 20, 2010, '0' RHR pump operability test (1ST);

• On April 25, 2010, Reactor coolant temperature during cool down per Technical Specification 4.6.A.2.;

• On April 28, 2010, 'A' main station battery performance test;

• On May 13,2010, MSIV 80C local leak rate test (CIV);

• On May 17, 2010, ECCS integrated automatic initiation test;

• On May 20, 2010, RPV operational system leakage test;

• On May 23, 2010, HPCI pump and valve operability test post refueling outage;

• On May 23, 2010, RCIC pump and valve operability test post refueling outage; and

• On June 28, 2010, reactor coolant system leak detection surveillance (RCS LO).

b. Findings

No findings were identified.

1EP6 Drill Evaluation

a. Inspection Scope

The inspectors observed an emergency preparedness (EP) drill on June 23, 2010, and reviewed the player and lead controller critiques. Entergy's EP staff preselected the drill notifications and protective action recommendations to be included in the EP drill performance indicator (PI). The inspectors discussed the performance expectations and results with Entergy's EP staff to confirm correct implementation of the PI program. The inspectors focused on the ability of licensed operators to perform event claSSifications, and to make proper notifications in accordance with Entergy's procedures and industry guidance. The inspectors evaluated the drill for conformance with the requirements of 10 CFR Part 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities." The inspectors compared Entergy's self-identified issues with observations from the inspectors' review to ensure that performance issues were properly identified and documented. The documents reviewed are listed in the

.

b. Findings

No findings were identified.

2RS0 1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

Radiological Hazard Assessment The inspector determined if there had been changes to plant operations since the last inspection that may result in a significantly new radiological hazard for onsite workers or members of the public. The inspector verified the licensee had assessed the potential impact of these changes with respect to the Spring refueling outage radiological conditions, and had implemented periodic monitoring, as appropriate, to detect and quantify the associated radiological hazards.

The inspector reviewed radiological surveys of principal refueling outage radiological work areas. The inspector reviewed surveys to verify they were appropriate for the given radiological hazards that were accessed by workers.

The inspector conducted walk-downs of the facility to evaluate material conditions and potential radiological conditions (radiological control area, protected area, controlled area, contaminated tool storage, and contaminated machine shops).

The inspector selected radiologically risk-significant work activities associated with Refueling Outage {RFO} 28 that involved exposure to radiation, including:

• Reactor in-service inspection;

• Reactor reassembly;

• Scaffold installation activities;

• Temporary shielding installation activities; and

• Radiation protection job coverage of various work activities.

The inspector reviewed pre-work surveys to verify if Entergy appropriately identified and quantified the radiological hazard, and established adequate protective measures. The inspector evaluated the radiological survey program to determine if hazards were properly identified,including the following:

• Identification of hot particles;

• The presence of alpha emitters;

• The potential for airborne radioactive materials, including the potential presence of transuranics and/or other hard-to-detect radioactive materials;

• The hazards associated with work activities that could suddenly and severely increase radiological conditions; and

• Severe radiation field dose gradients that can result in non~uniform exposures of the body.

The inspector selected air sample survey records to verify that samples were collected and counted in accordance with licensee procedures. The inspector observed work in potential airborne areas to verify that air samples were representative of the breathing air zone. The inspector verified that the licensee has a program for monitoring levels of loose surface contamination in areas of the plant with the potential for the contamination to become airborne.

Contamination and Radioactive Material Control At the Vermont Yankee Radiological Controlled Area (RCA) control point, the inspector observed workers surveying and releasing potentially contaminated materials for unrestricted use. The inspector verified that the counting instrumentation was located in a low background area and that the instruments sensitivity was appropriate for the type of contamination being measured.

Instructions to Workers The inspector selected containers holding nonexempt licensed radioactive materials resulting from the Spring 2010 refueling outage activities that may cause unplanned or inadvertent exposure of workers to verify that they were labeled and controlled.

The inspector reviewed radiation work permits (RWPs) associated with the work activities listed above that were used to access high radiation areas and identified what work control instructions or control barriers had been specified. The inspector verified that allowable stay times or permissible dose for radiologically Significant work under each radiation work permit was identified. The inspector verified that electronic personal dosimeter (EPD) alarm set points were in conformance with survey indications and plant policy.

The inspector selected two occurrences where a worker's EPD noticeably malfunctioned or alarmed. The inspector verified that workers responded to the off-normal condition.

The inspector verified that the issue was included in the CAP and dose evaluations were conducted.

This inspection effort represented partial completion of one sample. Further inspection is planned to fully complete this sample, and the results will be documented in a future report.

I.

I.

b. Findings

Ii No findings were identified.

2RS0 2 Occugational ALARA Planning and Controls

I

a. Inspection Scope

The inspector reviewed pertinent information regarding plant collective exposure history, current exposure trends, and ongoing or planned activities in order to assess current performance and exposure challenges. The inspector determined the plant's 3-year rolling average collective exposure.

The inspector determined the site-specific trends in collective exposures and source term measurements. The source term has been stable for the last several years with refueling outage dose rates as expected for most outage work activities.

This inspection effort represented partial completion of one sample. Further inspection is planned to fully complete this sample, and the results will be documented in a future report.

b. Findings

No findings were identified.

2RS0 4 Occupational Dose Assessment

a. Inspection Scope

Special Bioassa:t The inspector reviewed the adequacy of the licensee's program for dose assessments based on airborne/Derived Air Concentration (DAC) monitoring. The inspector verified that flow rates and/or collection times for fixed head air samplers or lapel breathing zone air samplers were adequate to ensure that appropriate lower limits of detection (LLDs}

are obtained. The inspector reviewed the adequacy of procedural guidance used to assess dose when the licensee applies protection factors. The inspector reviewed dose assessments performed using airborne/DAC monitoring. The inspector verified that the licensee's DAC calculations were representative of the actual airborne radionuclide mixture, including hard-to-detect nuclides.

The inspector reviewed the adequacy of the licensee's internal dose assessments for any actual internal exposure greater than 10 millirem committed effective dose equivalent. The inspector determined that the affected personnel were properly monitored with calibrated equipment and the data was analyzed and internal exposures properly assessed in accordance with licensee procedures.

This inspection effort represented partial completion of one sample. Further inspection is planned to fully complete this sample, and the results will be documented in a future report.

b. Findings

No findings were identined.

2RS05 Radiation Monitoring Instruments (71124.05 - 1 Sample)

a. Inspection Scope

During the period June 7-10, 2010, the inspector conducted the following activities to verify that the Vermont Yankee was ensuring the accuracy and operability of radiation monitoring instrumentation. Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, relevant Technical Specifications, and Vermont Yankee's procedures.

Inspection Planning

• The inspector reviewed the UFSAR to identify radiation instruments aSSOCiated with monitoring area radiological conditions including airborne radioactivity, process streams, effluents, material/articles. and workers;

• The inspector obtained a listing of all survey instrumentation including air samplers, small article monitors (SAMs), personnel contamination monitors (PCMs), and other monitors used to detect contamination. The inspectors reviewed the list to determine if an adequate number and type of instruments are available to support operations;

• The inspector obtained and reviewed copies of evaluation reports of the radiation monitoring program since the last inspection;

• The inspector obtained and reviewed copies of procedures used for instrument source checks and calibrations;

• The inspector reviewed area radiation monitor set point values and basis; and

• The inspector reviewed the effluent monitor set pOint basis and the calculation methods provided in the offsite dose calculation manual (ODCM).

Walkdowns and Observations

• The inspector toured the Turbine. Reactor, and Advanced Offgas buildings and observed the condition of the Reactor Building South Ventilation monitor, the Service Water Effluent monitor, and the Advanced Off Gas Hold Up and Stack monitors, and the Radioactive Liquid Waste Discharge monitor. These monitor configurations aligned with Vermont Yankee's ODCM descriptions;

• The inspector checked the calibration due dates and source check stickers for portable survey instruments ready for issue or in the field. The type of instruments checked included RO-20M, RM25. Telepoles, RM20, AMP-100, and REM-500;

• The inspector observed a technician perform instrument source checks. The inspectors verified that the instrument source checks included exposures at each high-range scale. The source check observations included RO-20M, RM20, Telepoles, AMP-100. and a Small Article Monitors (SAM)-11;

• . The inspector verified Area Radiation Monitors (ARM) and Continuous Air Monitors (CAM) were appropriately positioned relative to the radiation source(s) they were intended to monitor. The inspectors compared the monitor response with actual area conditions for severa! ARMs; and

• The inspector observed the daily source checks for Aptec AMP3e #204019. Thermo PM-7 #536 and 537, and SAM-11 #516 and 407. The inspector verified the source checks were in accordance with the manufacturer's recommendations and Vermont Yankee's procedures.

Calibration and Testing Program - Process and Effluent Monitors

• The inspector verified for fOLir effluent monitor instruments that channel calibration and functional tests were performed consistent with radiological effluent technical specifications. The inspector also verified that the source calibrations use National Institute of Standards and Technology (NIST) traceable sources or secondary measuring that has been calibrated to NIST standard. The inspectors verified that the sources used represent the plant nuclide mix;

• The inspector verified that effluent monitor alarm set pOints are established as provided in the ODCM and station procedures; and

• There were no changes to effluent monitor set-pOints during this inspection period.

Laboratocy Instrumentation

• The inspector verified that the daily performance checks and calibration data indicate the frequency of calibration is adequate and there is no degradation of instrument performance.

Whole Body Counter

• The inspector reviewed the methods and sources used to perform the Whole Body Counter (WBC) checks prior to daily use and observed a performance check. The inspector verified the checks are appropriate and align with the plant's isotopic mix; and

• The inspector reviewed the WBC calibration reports completed since the last inspection. The inspector verified the calibration sources and phantoms used were appropriate and representative of the plant source term.

Post*Accident Monitoring Instrumentation

• The inspector reviewed the May 16,2010 calibration records for the Drywell high range monitors, R16-181A and R16-18B; and three Containment Monitors. RM-18 154A, RM-18-154B, and RM-18-154C;

• The inspector verified that an electronic calibration for the Drywell high-range monitors was performed and included each decade above 10 rem/hour. The inspector also verified that a source calibration was performed and included an exposure for at least one decade below 10 rem/hour;

• The inspector verified the acceptance criteria were reasonable;

• The inspector reviewed the calibration records and availability for the Main Stack Ventilation and Reactor Building Ventilation high range monitors;

• The inspector reviewed Vermont Yankee's capability to collect high-range, post accident iodine effluent samples; and

• There were no opportunities to observe electronic or source calibrations of the high range monitors during this inspection. These calibrations are only performed during outages.

PMs. PCMs, and SAMs

• The inspector observed the use of an Eberline PM-7. PM-12, Aptec AMP3e, and SAM-11 and verified that the alarm set points were reasonable to ensure that licensed material is not released from the site; and

• The inspector reviewed calibration records and procedures for these devices and

.verified that calibration methodology was consistent with manufacturer's recommendations.

Portable Survey Instruments, ARMs. Electronic Dosimetry. and Air Samplers/CAMs

• The inspector reviewed calibration records for a PIG-6B, aTelepole, an R020AA, a Mark 2 ElectroniC Personal Dosimeter, an AMS-4, an RP-4520. and a REM-500.

The inspector reviewed the detector measurement geometry and calibration methods for ARMs and portable radiation survey instruments. The inspector observed a calibration of a PIG-6B and a source check of a Telepole; and

• The inspector reviewed the corrective actions taken for two R020s and one Telepole found to be out of the acceptance criteria during the observed daily source check.

1\10 corrective action reports were identified for personal monitoring devices.

Instrument Calibrator

• The inspector reviewed the current output tables for Vermont Yankee's portable survey and ARM instrument calibrator unit. The inspector verified that Vermont Yankee periodically measures the calibrator output over the range of the instruments; and

• The inspector verified the calibrator is periodically calibrated using NIST traceable sources.

Calibration and Check Sources

• The inspector reviewed Vermont Yankee's 10 CFR 61 source term to verify that the calibration sources used are representative of the types and energies of radiation encountered in the plant.

Problem Identification and Resolution

• The inspector reviewed condition reports related to radiation monitoring instrumentation and selected three specific CRs to review with Vermont Yankee staff. The inspector verified that appropriate corrective actions have been taken.

The inspector verified that problems are being identified at the appropriate threshold and are properly addressed for resolUtion.

This inspection effort represented completion of one sample.

b. Findings

No findings were identified.

OTHER ACTIVITIES

lOA]

40A1 Performance Indicator (PI) Verification (71151 - 2 samples)

Initiating Events Cornerstone a.

Inspection ScoRe The inspectors reviewed Entergy's submittals and performance indicator (PI) data for the cornerstones listed below for the period from January 2009 to December 2009. The inspectors reviewed selected operator logs, plant process computer data, licensee event reports, and CRs. The PI definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Performance Indicator Guideline," and AP 0094, "NRC Performance Indicator Reporting," were used to verify the accuracy and completeness of the PI data reported during this period. The Pis reviewed were:

• Reactor coolant system activity (BI01); and

• Reactor coolant system total leakage (BI02).

b. Findings

No findings were identified.

40A2 Identification and Resolution of Problems (71152 - 1 sample)

.1 Reviews of Items Entered into the Corrective Action Program

a. Inspection Scope

The inspectors performed a daily screening of each item entered into Entergy's CAP.

This review was accomplished by reviewing printouts of each CR, attending daily screening meetings, and/or accessing Entergy's 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.

Additionally, the inspectors reviewed a sample of non-conforming conditions identified during lSI examinations this outage to evaluate the effectiveness of the licensee in identification and resolution of problems. The inspectors selected INR VYR28-3 RO and R2 for evaluation of flaw identification, characterization and placement into the licensee corrective action program using CR-VTY-2010-02016. The remote in-vessel, visual inspection (lWI) of the reactor steam dryer welds revealed crack-like indications at the DC-V4C Steam Dryer Drain Channel weld. This flaw was previously identified during visual examination in refuel outage RFO-27. The indication was characterized as linear, located in the heat affected zone on the drain pipe DC-H-27 side of the weld. The vertical component of this indication does not connect to the previously identified indication on drain channel H-27. The inspector reviewed the materials of construction, weld metal, flaw location, weld metal profile and condition of the adjacent base material.

The inspector verified that the flaws identified were characterized appropriately and entered into the licensee's corrective action process.

b. Findings

No findings were identified .

.2 Semi-Annual Trend Review (1 sample)

a. Inspection ScoQe The inspectors performed a review of Entergy's CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors' review considered the six~month period of November 2009, to April 2010, although some examples expanded beyond those dates when the scope of the trend warranted. The inspectors compared their results with the results contained in Entergy's quarterly trend reports, operator logs, and CRs. The corrective actions assigned to address select individual issues were reviewed for adequacy.

b. Findings and Observations

No findings or observations of significance were identified.

40A3 Event Followup (71153 - 3 samples)

.1 Reactor Scram

a. Inspection Scope

At 3:29 p.m. on May 25, 2010, the Vermont Yankee nuclear power plant experienced a main generator trip and lockout due to a high differential current on a 345KV Tie Line.

This resulted in a main turbine trip and reactor scram. The plant did not experience any complications during the scram and all automatic functions responded appropriately.

Prior to the main generator trip, the licensee was raising reactor power from 70 percent to 74.5 percent at 1 percent every 3 minutes. This is the highest power level reached after tying in the new VELCO switchyard to 345KV system, Entergy determined that the main generator trip was initiated by a high differential current caused by differences in the winding ratios between the current sensors in the VY switchyard and the new VELCO switchyard. As the generator power was raised, the current sensors deviated sufficiently to cause the main generator to trip. Entergy generated a modification change to make Entergy's current sensors compatible with VELCO's current sensors.

b. Findings and Observations

No findings or observations of significance were identified .

.2 Loss of Communications

a. Inspection Scope

At 2:00 a,m. on May 27.2010, Vermont Yankee lost telephone communications. The loss of telephone communications was due to the high winds experienced during a thunderstorm. Lost external communication capability inCluded commercial telephone and the Emergency Notification System. A microwave link with the local transmission system operator and some cell phone service remained available. At the time of the loss of communications, the plant was shutdown following a scram caused by unrelated issues. An emergency (Notice of Unusual Event) was not declared because sufficient internal and external communication capability remained available. Communication capabilities were restored at approximately 2:00 p.m.

b. Findings and Observations

No findings or observations of Significance were identified .

.3 Notice of Unusual Event

a. Inspection Scope

On June 23, 2010. Vermont Yankee declared a Notice of Unusual Event (UE) at 2:25 p.m. The declaration was made by Entergy in accordance with the plant specific emergency action guidelines after ground motion was felt by personnel on-site and the occurrence of an earthquake was confirmed by the National Earthquake Information Center. According to U.S. Geological Survey, an earthquake of magnitude 5.0 occurred with a seismic epicenter located 33 miles north of Gatineau, Quebec, Canada, near Ottawa and approximately 300 miles from the Vermont Yankee site. Vermont Yankee continued to operate normally at 100-percent power and there was no evidence of damage based on plant inspection of station equipment and structures. The NRC Resident Inspectors were on-site and independently verified, by conducting inspection activities, that there was no evidence of damage to plant equipment. The NRC confirmed that VY operators notified the states of NH, VT, and MA regarding the UE emergency declaration. The UE was terminated by Entergy at 5:25 p.m. on June 23, 2.01.0, based on a thorough investigation that verified no damage to station eqUipment and structures had occurred as a result of the seismic activity.

b. Findings and Observations

No findings or observations of significance were identified.

40A5 Other Activities TI 2515/179 Verification of Licensee Resl20nses to NRC Requirement for Inventories of Materials Tracked in the National Source Tracking System Pursuant to Title 10. Code of Federal Regulations, Part 2.0.2207 (1

.0 CFR 2.0.22.07)

a. Inspection Scope

During the period June 7, 2.010, through June 10, 201.0, the inspector conducted the following activities to confirm the inventories of materials possessed by Vermont Yankee were appropriately reported and documented in the National Source Tracking System (NSTS) in accordance with 10 CFR 20.2207

Inspection Planning

• The inspector retrieved a copy of Vermont Yankee's NSTS inventory. The inventory indicated that Vermont Yankee didn't possess any sources that required NSTS tracking.

Inventory Verification

• The inspector reviewed the site's source inventory and verified aU sources were below the level requiring NSTS tracking.

• The inspector discussed the requirements with Vermont Yankee staff members, including the Radiological Protection Support Supervisor, and determined that they were aware of the requirements and had correctly determined that they didn't apply to Vermont Yankee at this time. However, Vermont Yankee understood the need to comply with the NSTS requirements if applicable sources were obtained in the future.

b. Findings

No findings were identified.

40A6 Meetings, including Exit

Exit Meeting Summary

On May 13. 2010, the inspector presented the radiation safety baseline inspection results to Mr. Christopher Wamser, General Manager of Plant Operations, and other members of his staff. The inspector confirmed that no proprietary information was provided or examined during the inspection.

On June 10, 2010, the inspector presented the radiation safety and TI 179 inspection results to Mr. David Tkatch, Radiation Protection Manager. The inspector verified that proprietary information was not provided to the inspector for the inspection.

On June 21, 2010, Donald Jackson, NRC Regional Branch Chief for Vermont Yankee, presented and discussed the End-of-Cycle performance assessment of the Vermont Yankee Nuclear Power Plant with Mr. Michael Colomb, Site Vice President. The licensee acknowledged the assessment and planned regulatory oversight. This discussion was completed prior to a public performance assessment open-house and meeting on June 22,2010. (ADAMS Accession ML100621409).

nd On July 26, 2010, the resident inspectors presented the 2 quarter inspection results to Mr. Michael Colomb, Site Vice President. and other members of the Vermont Yankee staff. The inspectors confirmed that no proprietary information was provided or examined during the inspection.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Vermont Yankee Personnel

M. Colomb, Site Vice President

C. Wamser, General Manager of Plant Operations

G. Lozier, Acting Director of Nuclear Safety

J. Devincentis, Acting Licensing Manager

N. Rademacher, Director of Engineering

M. Philippon. Operations Manager

J. Rogers, Design Engineering

P. Rose, Operations/FIN Team

G. Von der Esch, Asst. Operations Manager

R. Current, Sr. Electrical I&C System Engineer

L. Doucette, System Engineering

R. Meister, Licensing

P. Corbett. Manager, Quality Assurance

P. Couture, Licensing Specialist

L. Derting, Supervisor, Radwaste

J. Geyster, Supervisor, Radiation Protection

M. Gosekamp, Manager, Maintenance

J. Hardy, Superintendant, Chemistry

R. Wanczyk, Acting Manager, CA&A

M. Morgan, Superintendent, Training

S. Skibniowski, Environmental Specialist

P. Stover, Supervisor, Radiation Protection

D. Tkatch, Manager, Radiation Protection

R. Wanczyk, Enexus Site Representative

K. Stupak, LOR Program Lead

H. Sreite, Senior engineer

J. Mully, Engineer

S. Jonasch, Engineer

R. Current, Engineer

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000271/2010003-01 NCV Inadvertent Loss of RCS Inventory During ECCS Testing Due to Inadequate Procedure

Closed

None

Discussed

None