IR 05000443/2008003

IR 05000443-08-003; 04/01/2008 - 06/30/2008; Seabrook Station, Unit No. 1; Outage Activities and Access to Radiological Significant Areas

ML082140855
Person / Time
Site:	Seabrook
Issue date:	08/01/2008
From:	Arthur Burritt Reactor Projects Branch 3
To:	O'Keefe M, St.Pierre G Florida Power & Light Energy Seabrook
Burritt A RGN-I/DRP/PB3/610-337-5069
References
EA-08-164 IR-08-003
Download: ML082140855 (58)

Text

UNITED STATES ust 1, 2008

SUBJECT:

SEABROOK STATION, UNIT NO. 1 - NRC INTEGRATED INSPECTION REPORT 05000443/2008003

Dear Mr. St. Pierre:

On June 30, 2008, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at the Seabrook Station, Unit No. 1. The enclosed report documents the inspection findings discussed on July 1, 2008, with Mr. G. S and other 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 three self-revealing findings of very low safety significance (Green) that were determined to involve violations of NRC requirements. However, because of their very low safety significance and because they were entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs), in accordance with Section VI.A.1 of the NRC Enforcement Policy.

Additionally, a licensee-identified violation that was determined to be of very low safety significance is listed in this report.

If you contest any NCV in this report, 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 Seabrook Station.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and 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

NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Arthur L. Burritt, Chief Projects Branch 3 Division of Reactor Projects Docket No. 50-443 License No: NPF-86

Enclosure:

Inspection Report No. 05000443/2008003 w/ Attachment: Supplemental Information

REGION I==

Docket No.: 50-443 License No.: NPF-86 Report No.: 05000443/2008003 Licensee: FPL Energy Seabrook, LLC (FPLE)

Facility: Seabrook Station, Unit No. 1 Location: Seabrook, New Hampshire 03874 Dates: April 1, 2008 through June 30, 2008 Inspectors: William Raymond, Senior Resident Inspector J. Johnson, Resident Inspector R. Moore, Project Engineer L. Scholl, (Acting) Resident Inspector D. Silk, (Acting) Resident Inspector G. Johnson, (Acting) Resident Inspector T. Moslak, Health Physicist T. Burns, Reactor Inspector A. Ziedonis, Reactor Inspector Approved by: Arthur L. Burritt, Chief Projects Branch 3 Division of Reactor Projects

SUMMARY OF FINDINGS

IR 05000443/2008003; 04/01/2008 - 06/30/2008; Seabrook Station, Unit No. 1; Outage

Activities and Access to Radiological Significant Areas.

The report covered a three-month period of inspection by resident inspectors, a regional reactor inspector, and an announced inspection by a regional health physics specialist. Three Green non-cited violations (NCVs) were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or 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 non-cited violation of Technical Specification 6.7.1.a was identified for the failure to implement written procedures governing safety-related activities. Specifically, on April 20, 2008, FPL Energy Seabrook (FPLE) failed to implement tagging and configuration control procedures, resulting in the loss of configuration control during shutdown operations when flow was established through a partially disassembled charging system valve. This resulted in a 200 gallon leak of reactor cavity water onto the floor of the Primary Auxiliary Building (PAB). The letdown flow path was established while work was in progress on valve CS-V-299. A clearance boundary was modified with the incorrect assumption that CS-V-299 was intact.

This finding was more than minor because it was associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of plant events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the loss of configuration control in the charging system unintentionally drained 200 gallons from the reactor cavity, which affected the shutdown critical safety function of maintaining adequate reactor inventory, and caused an uncontrolled leak of radioactively contaminated water to a work area. The finding was determined to be of very low safety significance (Green)using the SDP Appendix G assessment, since the finding did not result in a loss of control of shutdown operations and adequate mitigation capabilities remained available. FPLE entered this issue into the corrective action program as Condition Report 08-06270.

The finding has a cross-cutting aspect in the area of human performance, work control, since FPL Energy did not plan and coordinate work activities consistent with nuclear safety (H.3(b)). Specifically, FPLE revised a clearance tagging boundary without verifying the status of affected work activities in accordance with site procedures. (Section 1R20)

Green.

A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified because FPLE did not implement corrective actions to prevent recurrence of mispositioned valves caused by difficult to operate stow-operator reach rods. Specifically, on April 20, 2008, a mispositioned (partially open), stow-operated filter drain valve, CS-V-1190, resulted in the inadvertent draining of 2000 gallons of water from the reactor cavity while operators placed the reactor letdown system into service. The drain valve was partially open because it was difficult to operate when positioned with its stow-operator. The mispositioning of a stow-operated valve in a safety system was a repeat occurrence of a similar event in October 2007.

This finding was more than minor because it was associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of plant events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the loss of configuration control in the charging system unintentionally drained 2000 gallons from the reactor cavity, which affected the shutdown critical safety function of maintaining adequate reactor inventory. The finding was determined to be of very low safety significance (Green) using the SDP Phase 1 assessment, since the finding did not result in a loss of control of shutdown operations and adequate mitigation capabilities remained available. FPLE entered this issue into the corrective action program as Condition Report 08-06260.

The finding has a cross-cutting aspect in the area of problem identification and resolution because FPL Energy did not take appropriate corrective actions to address safety issues in a timely manner commensurate with their safety significance and complexity (P.1.d). Specifically FPL Energy did not take adequate corrective actions to assure the correct positioning of stow-operated safety system valves and thereby prevent recurrence of a significant condition adverse to quality. (Section 1R20)

Cornerstone: Occupational Radiation Safety

Green.

A self-revealing non cited violation of Technical Specification 6.11.2 was identified. Specifically, on May 1, 2008, FPLE failed to identify and control an existing high radiation area with dose rates greater than 1000 millirems per hour in the reactor containment building. A worker was exposed to higher than expected radiation levels of approximately 2,270 mrems per hour. The worker received a dose of 4 millirem.

The finding is more than minor because it is associated with the occupational radiation safety cornerstone attribute of exposure control and affected the cornerstone objective, because not controlling the locked high radiation areas could increase personal exposure. The finding was determined to be of very low safety significance (Green) using the SDP assessment because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. FPLE entered the issue into the corrective action program as a Condition Report 200806982.

This finding had a cross-cutting aspect in the area of human performance, work control, because FPLE did not appropriately plan work by incorporating job site conditions that impact radiological safety. Specifically, FPLE did not adequately assess changing area dose rates caused by operating activities, and thus did not adequately plan a work task with due consideration of the actual radiological conditions at the job site (H.3(a)). (2OS1).

Licensee-Identified Violations

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

REPORT DETAILS

Summary of Plant Status

Seabrook, Unit No. 1 (Seabrook) operated continuously at or near full power for the duration of the inspection period except for a planned refueling outage that began on April 1, 2008, and completed on May 8, 2008. FPL Energy (FPLE) completed refueling, testing and maintenance activities during the outage. This included loading new fuel in the reactor, placed overlay welds on six pressurizer nozzles, modified the containment sump, and replaced components in the 345KV electrical switchyard. FPLE also completed a containment integrated leak rate test.

Seabrook returned to 100 percent power on May 11, 2008, and remained at full power until June 5, when power was reduced to 30% FP due to a condenser tube leak. Full power operations resumed on June 8 and continued for the remainder of the period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Preparation

.1 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspector completed one seasonal extreme weather conditions inspection sample.

The inspectors assessed FPLEs readiness for the onset of extreme hot weather conditions. The inspectors reviewed the updated final safety analysis report (UFSAR)descriptions for related design features and verified the adequacy of the station procedures for hot weather protection. The inspectors reviewed FPLEs actions per procedure ON1490.09 for seasonal readiness, and procedure OS1200.03 for severe weather. The inspectors also conducted walkdowns of susceptible systems, specifically the emergency feedwater and service water systems. The inspectors reviewed deficiencies related to extreme weather preparation and verified the issues were entered into the corrective action program. The references used for this review are listed in A.

b. Findings

No findings of significance were identified.

.2 Summer Readiness of Offsite and Alternate AC Power Systems

a. Inspection Scope

The inspectors completed one summer readiness of offsite and alternate AC power systems inspection sample. The inspectors review of this area focused on FPLE procedure OS1246.02, Degraded Vital AC Power. The inspectors verified that plant features were maintained and procedures for operation were adequate to ensure the continued availability of AC power systems. The inspectors verified that communication protocols with the transmission system operator were adequate to ensure that appropriate information was exchanged when issues arose that could impact the offsite power system. The inspectors also observed FPLEs implementation of OS1246.02 during periods of high ambient temperatures that occurred between June 7 and 10, 2008. The inspection included walkdowns of the onsite normal and emergency AC power systems and the inspectors reviewed deficiencies related to summer readiness of offsite and alternate AC power systems and verified these issues were entered into the corrective action program. The references used for this review are listed in A.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment (71111.04 - 4 samples,

==71111.04S - 1 sample)

.1 Partial System Walkdown

==

The inspectors performed a partial system walkdown on the four plant systems listed below. The inspectors completed walkdowns to determine whether there were discrepancies that could impact the function of the system, and therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, walked down control system components, and verified that selected breakers, valves, and support equipment were in the correct position to support system operation. The inspectors also verified that FPLE had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program. The references used for this review are listed in Attachment A.

The inspectors performed the following partial system walkdowns:

• Reactor vessel level instrumentation for shutdown operations on April 4-8;

• Boration flow path from the RWST to safety injection pump 6A with injection into the RCS cold legs on April 10-14;

• Service water system lineup to cooling tower train B during service water system maintenance on April 10-17; and

• A EDG during maintenance on the B EDG on June 16 - 20.

b. Findings

No findings of significance were identified.

.2 Complete System Walkdown

a. Inspection Scope

The inspectors performed a complete system walkdown inspection of the B Loop of the residual heat removal system to verify the system was properly aligned and capable of performing its safety function. To ascertain the required system configuration, the inspectors reviewed plant procedures, system drawings, the UFSAR, and the TS. The references used for this review are listed in Attachment A. The inspectors walked down the accessible portions to the system to verify the systems overall material condition; that valves were correctly positioned; that electrical power was available; that major system components were properly labeled; that hangers and supports were correctly installed and functional; and that ancillary equipment or debris did not interfere with system performance.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Quarterly Review of Fire Areas

a. Inspection Scope

The inspectors completed seven quarterly fire protection inspection samples. The inspectors examined several areas of the plant to assess: the control of transient combustibles and ignition sources; the operational status and material condition of the fire detection, fire suppression, and manual fire fighting equipment; the material condition of the passive fire protection features (fire doors, fire dampers, fire penetration seals, etc.); and the compensatory measures for out-of-service or degraded fire protection equipment. The following areas were inspected:

• Containment Building 26 Area (Zone C-F-1-Z)

• Containment Building 0 Area (Zone C-F-2-Z)

• Containment Building (-)26 Area (Zone C-F-3-Z)

• Electrical Tunnel Train B (Zones ET-F-1C)

• A Train Diesel Generator Room 21 Area (Zone DG-F-2A-A)

• B Train Diesel Generator Room 21 Area (Zone DG-F-2B-A)

• Emergency Feed Water Pumphouse 27 Area (Zone EWP-F-1-A)

The inspectors verified that the fire areas were maintained in accordance with applicable portions of Fire Protection Pre-Fire Strategies and Fire Hazard

Analysis.

NRC review of this area is also discussed in Section 4OA3 of this report. The inspector periodically toured the containment during the refueling outage to review FPLE controls of transient combustibles and reviewed FPLE actions to address deficiencies in the corrective action program. The references used for this review are listed in Attachment A.

b. Findings

No findings of significance were identified.

.2 Annual Inspection

a. Inspection Scope

The inspectors completed one annual sample to evaluate fire brigade performance. The inspectors observed an announced fire brigade drill on June 10, 2008, on the 53 elevation of the PAB. The inspectors observed brigade performance during the drill to evaluate the following: donning and use of protective equipment; fire brigade leader command and control; fire brigade response time; radio communications; and the use of pre-fire plans. The inspectors attended the post-drill critique and reviewed the disposition of issues and deficiencies identified during the drill. The inspectors also verified that all fire fighting equipment used during the drill was returned to a condition of readiness. This review covered one inspection sample. The references used for this review are listed in Attachment A.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

.1 Annual Inspection

a. Inspection Scope

The inspectors reviewed Seabrooks program for monitoring the B primary component water heat exchanger CW-E-17B to determine whether the heat exchanger could fulfill its design function. The inspectors reviewed past thermal performance monitoring, trending data for heat exchanger temperatures and fouling factors, and ES1850.017, "SW Heat Exchanger Program," Revision 0. The inspectors reviewed data monitored by the system engineer to evaluate the process used to monitor the heat exchanger and commitments in Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." The inspectors also reviewed condition reports to verify that heat exchanger thermal performance issues were identified and corrected, including condition report (CR) 200805414 and work order WO 0641218. The references used for this inspection are listed in Attachment A.

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection

a. Inspection Scope

The purpose of this inspection was to review and assess the effectiveness of Seabrooks Inservice Inspection (ISI) program for monitoring degradation of the reactor coolant system (RCS) boundary, risk significant piping system boundaries, and the containment boundary. The inspectors reviewed the inservice inspection activities using the criteria specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI and applicable NRC Regulatory Requirements.

The inspectors selected a sample of nondestructive examination (NDE) activities to review for compliance with the requirements of ASME Section XI. Also, the inspectors selected samples of modification, repair and replacement activities that involved use of the welding process on pressure boundary risk significant systems. The sample selection was based on the inspection procedure objectives, risk significance and availability. The inspectors reviewed examination procedures, procedure and personnel qualifications and examination test results. The inspectors reviewed samples of examination reports and condition reports (CR) initiated during ISI examinations to evaluate FPLEs effectiveness in the identification and resolution of problems.

The inspectors reviewed the procedures used to perform visual examinations for indications of boric acid leaks from pressure retaining components, including the vessel upper head penetrations and their connections to the drive mechanisms. The inspectors reviewed a sample of condition reports initiated as a result of the inspections performed in accordance with FPLEs boric acid control program. The inspectors selected CRs that identified both active and inactive leak locations that could result in degradation of safety significant components. The inspectors reviewed six CRs shown on Attachment A that identified boric acid crystal deposits identified during plant walkdowns performed during and after plant shutdown. The inspectors reviewed samples of operability evaluations, engineering evaluations and corrective actions provided for active and inactive boric acid leaks and determined they were consistent with the requirements of the ASME Code and 10 CFR 50, Appendix B, Criterion XVI.

The inspectors observed the performance of two NDE activities in process and reviewed documentation and examination reports for an additional four nondestructive examinations. Non-destructive test processes included, visual (VT), magnetic particle (MT), penetrant (PT) and ultrasonic (UT) testing.

ISI examinations that were reviewed included:

• Ultrasonic test of weld RC-E-10-A-SWOL overlay deposited on pressurizer A safety relief nozzle to safe end dissimilar metal weld, drawing 8020570C Rev 0.

• Liquid penetrant test of weld RH 0179-01 01, butt weld of pipe to T, residual heat removal (RHR) system, drawing 1-NHY-800179 ISI.

• Magnetic particle test of weld MS 4002-02 09, butt weld of elbow to pipe, main steam (MS) system, drawing 1-NHY-202303 ISI.

• Magnetic particle test of weld FW 4606-03 06, butt weld of valve V-30 to pipe, feedwater (FW) system, drawing 1-NHY-202396 ISI.

The inspectors performed a walk-down of portions of the containment liner on the zero

(0) twenty five

(25) and minus twenty six (-26) foot elevations to inspect the condition of the coating on the primary containment liner per ASME Section XI Section IWE. The inspectors also inspected examination reports of the results of FPLEs examination. In addition, the inspectors interviewed the containment liner program manager to determine the scope of containment boundary examinations and management oversight of the activity during this outage.

The inspectors reviewed the steam generator (SG) condition monitoring assessment and operational assessment to evaluate FPLEs conclusion that no SG tube inspection was required for this outage. The inspector noted FPLEs technical evaluation and determination that there were no degradation mechanisms in the Seabrook SGs that are ongoing or active and that all structural criteria will be satisfied until the next scheduled refuel outage (13).

The inspectors reviewed documentation for two rework/repair activities that required the development of an ASME Section XI repair plan with the use of welding processes to complete the repair. The work orders (WOs) governing these repair activities were:

• WO 0615086, Addition of vent with isolation valve on containment building spray, train A, safety-related, Code Class 2, ASME Section III.

• WO 0643870, Installation of suction and overflow piping from TK-RWST-8 and installation of piping and supports downstream of valve 1-CBS-V35.

The inspectors also reviewed the ASME Section XI repair plans, replacement material, weld procedure specifications and qualifications, welder qualifications, weld filler metals, non-destructive tests, acceptance criteria and post work testing for each activity, as applicable.

The inspectors reviewed a sample of condition reports (CR) related to inservice inspection activities. The specific CRs reviewed and listed in Attachment A. The inspectors determined that the nonconforming conditions identified were reported, characterized, evaluated and appropriately dispositioned and entered into the corrective action program.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

.1 Quarterly Resident Inspector Review

a. Inspection Scope

The inspectors completed one licensed operator requalification program quarterly inspection sample. The inspectors observed the conduct of licensed operators during simulator training sessions on May 19 and 22, 2008, the details of which are described below. The references used for this inspection are listed in Attachment A

• On May 19, the inspectors observed a simulator training session to review operator actions to implement procedures OS1290.03 and OS1290.04. The inspectors reviewed the operators actions to implement plant emergency procedures, classify events under the emergency plan and coordinate emergency actions with other response organizations.

• On May 22, the inspectors observed a simulator training session to review operator actions to implement the emergency operating procedures. The inspectors reviewed the simulators physical fidelity in order to verify similarities between the Seabrook control room and the simulator. The inspectors examined the operators ability to perform actions associated with high-risk activities, the Emergency Plan, previous lessons learned items, and the correct use and implementation of procedures. The inspectors observed and reviewed the training evaluators critique of operator performance and verified that deficiencies were adequately identified, discussed, and entered into the corrective action program, as needed.

b. Findings

No findings of significance were identified

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors completed two maintenance effectiveness quarterly inspection samples.

The samples included one system review and one specific issue review. The inspectors evaluated maintenance rule implementation for the solid state protection system and the heater drain system. The inspectors reviewed the effectiveness of maintenance through a review of deficiencies identified, historical performance, and overall system performance. The inspectors also reviewed the Seabrook UFSAR and TS for these systems and examined maintenance rule functional failure (MRFF) evaluations against the guidance in NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Rev. 2. Other references used for this inspection are listed in Attachment A.

For the solid state protection system the inspectors assessed: 1) the application for MR scoping and MR reliability/availability performance criteria; 2) the corrective actions for deficient conditions; 3) the extent-of-condition reviews for common cause issues; and 4)the contribution of deficient work controls or work practices to any degraded conditions.

FPLE corrective actions were assessed against 10 CFR 50.65 requirements and the guidance in NUMARC 93-01. The inspectors interviewed licensee personnel; reviewed condition reports, procedures, and photographs; and observed activities regarding the discovery, trouble-shooting, and resolution of a problem associated with over voltage protection devices (OVPD) for the new power supplies to the solid state protection system. The inspectors also reviewed FPLEs extent-of-condition assessment regarding the OVPDs.

For the heater drain system the inspectors reviewed maintenance rule functional failure determinations associated with failures that occurred between 2006 and 2008. The inspector also reviewed the corrective actions for each failure.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

The inspectors reviewed the scheduling and control of seven emergent work troubleshooting activities to evaluate the overall effect on plant risk. The inspectors conducted interviews with operators, risk analysts, maintenance technicians, and engineers to assess their knowledge of the risk associated with the work, and to ensure that appropriate risk management actions were implemented. The actions taken were evaluated using the following Seabrook procedures: Maintenance Manual 4.14, "Troubleshooting, Revision 0 and Work Management Manual 10.1, "On-Line Maintenance, Revision 3. Specific risk assessments were conducted using Seabrook's "Safety Monitor." The inspectors reviewed the following emergent work activities:

• Reactor makeup water (RMW) Valve Seat Leakage (WO 0817973, CR200807918)

• Repair of Main Steam Vent Valve 1-MS-V298 by Leak Seal Process (WO 0817619)

• Repair of 345KV Switchyard Bus 2 fault - zone isolation valve (WO 0815298)

• Chemical Volume and Control System dilution (WO 0817656)

• Repair of heater drain pump HD-P-31A (WO 0817798)

• Repair of a Condenser Tube Leak in CO-E-27C (WO 0818508)

• Component cooling heat exchanger CC-E17A Fouling (WO 0641218, 0814321).

The inspectors interviewed engineering personnel and reviewed photographs and condition reports regarding grass that was found inside the B PCCW heat exchangers. The inspectors reviewed the licensees extent-of-condition assessment for impact on other heat exchangers cooled by service water.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors completed five operability evaluation inspection samples. The inspectors reviewed operability evaluations and/or condition reports to verify that the identified conditions did not adversely affect safety system operability or overall plant safety. The evaluations were reviewed using criteria specified 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 and Inspection Manual Part 9900, "Operability Determinations and Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety." In addition, where a component was determined to be inoperable, the inspectors verified that any TS limiting condition for operation implications were properly addressed. The inspectors also performed field walkdowns, interviewed personnel involved in identifying, evaluating or correcting the identified conditions. The following five items were reviewed:

• CR200803566 that addressed ECCS System Operability in Mode 4;

• CR200804857 that addressed the adequacy of the pressurizer vent opening for overpressure protection;

• CR200805414 that addressed the fouling of the primary component cooling water heat exchangers;

• CR200808673 that addressed the fuel consumption rate for the A EDG; and

• CR200808703 that addressed leakage from SW-V-92.

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

.1 Permanent Modification - Containment Sump Strainers

a. Inspection Scope

The inspectors completed one plant modifications inspection sample. The inspectors reviewed the design changes associated with the modification of the containment sumps performed under 06DCR008. Modification 06DCR008 implemented changes to the sumps as part of the actions to resolve licensee commitments under Generic Letter 2004-02. The modification replaced the existing sump screens with screens that have larger surface area and include fine mesh for debris removal. The inspectors reviewed the changes made to the existing structures and the engineering and design bases supporting the modification. The inspectors interviewed engineers and project staff.

The inspectors reviewed FPLEs safety evaluation screening for the modification completed per the requirements of 10 CFR 50.59. The inspectors also walked down the strainer fabrication and installation areas to verify compliance with the design documents. The inspectors reviewed the post-modification closure of the sumps and containment to ensure they were appropriate to support plant operations. Section 4OA5.2 of this report also describes additional NRC reviews that were completed in this area. The references used for this review are listed in Attachment A.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors completed six post-maintenance testing inspection samples. The inspectors reviewed post-maintenance testing (PMT) activities to ensure: the PMT was appropriate for the scope of the maintenance work completed and in accordance with MA 3.5, "Post Maintenance Testing;" the acceptance criteria were clear and demonstrated operability of the component; and the PMT was performed in accordance with procedures. The inspectors review the PMT for the following maintenance activities:

• WOs 071390 and 0633928 that included an oil leak repair, and MOV inspection and spring pack replacement for the safety injection cold shutdown test valve, SI-V-77

• WO 0640186 that was written to address an oil sample containing debris, and included an inspection and repair of the motor inboard bearing for the B charging pump, CS-P-2B

• WO 0706819 that specified completion of reserve auxiliary transformer preventive maintenance and doble testing

• WO 0639798 that tested the solid state system power supplies after modification and repairs

• WOs 0733190, 0619300, and 0619299 that tested the A charger for the 125 Vdc battery following repairs

• WOs 0821178, 0821176, 0717730, and 0803491 that completed testing for the B emergency diesel generator following repairs.

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities

a. Inspection Scope

The inspectors reviewed the operational, maintenance, and testing activities for the twelfth refueling outage OR12 starting on April 1, 2008. The references used to this inspection are listed in Attachment A.

Review of Outage Plan The inspectors reviewed the outage plans to evaluate Seabrooks ability to assess and manage the outage risk. The inspectors reviewed the outage risk assessment provided in Engineering Evaluation EE-08-004.

Monitoring of Plant Shutdown and Cooldown Activities The inspectors reviewed FPLE action to shut the plant down in accordance with plant procedures. The inspectors observed completion of various activities required to place the plant in a cold shutdown condition to assess operator performance, communications, command and control and procedure adherence. The inspectors reviewed operator adherence to TS required cooldown limits.

The inspectors also conducted inspection walkdowns of plant areas not normally accessible during plant power operations to verify the integrity of structures, piping and supports, and to confirm that systems appeared functional.

Core Reload Fuel Shuffle Activities and Reactivity Control The inspectors verified that refueling activities were conducted in accordance with procedures OS1000.09 and RS0721. The inspectors independently verified on a sampling basis that requirements for core alteration were met. The inspectors observed FPLE actions during core alterations to assure core reactivity was controlled. The inspectors observed activities from the control room, the reactor cavity and the spent fuel pool at various times. The inspectors verified that fuel movement was tracked in accordance with the fuel movement schedule. The inspectors verified FPLE action to meet the requirements of TS 3.9 for refueling operations, including the requirements for boron concentration and core monitoring using the source range monitors. The inspectors observed communications and coordination of activities between the control room and the refueling stations while fuel handling activities were in progress.

Outage Risk The inspectors reviewed daily shutdown risk assessments during refueling outage OR-12 to verify that FPLE addressed the outage impact on defense-in-depth for the critical safety functions: electrical power availability, inventory control, decay heat removal, reactivity control, and containment. The inspectors reviewed how FPLE provided adequate defense-in-depth for each safety function, and implemented the planned contingencies in order to minimize the overall risk where redundancy was limited or not available. The inspectors periodically reviewed risk updates accounting for schedule changes and unplanned activities.

Control of Heavy Loads The inspectors reviewed FPLEs activities to control the lift of heavy loads in accordance with plant procedures and the commitments to NUREG 0612. The inspectors observed the preparations for and in-progress lift activities to verify adherence to established procedures and controls. The inspectors used operating experience smart sample OpESS 2007-03 as a reference for this review. The inspection included a review of the updated design and licensing basis, as described below.

The inspectors reviewed FPLEs actions to implement the controls described in Nuclear Energy Institute (NEI) Letter, Guidelines for Reactor Vessel Head Drop Analyses, dated January 15, 2008. FPLE revised the Seabrook design basis by completing a reactor head drop analysis as part of the controls for handling heavy loads. The inspectors reviewed FPLEs actions to implement safe load paths, implement load handling procedures, use qualified crane operators, use special lifting devices and complete inspection, testing and maintenance of cranes. The inspectors reviewed the load drop analysis and verified that the analysis bounded the planned lifts with respect to load weight, load height, medium under the load and procedures that implement the safety analysis. FPLE actions to update UFSAR to reflect the new design basis were in progress at the end of the inspection period.

Clearance Activities and Configuration Control The inspectors reviewed a sample of risk significant clearance activities and verified tags were properly hung and/or removed, equipment was appropriately configured per the clearance requirement, and that the clearance did not impact equipment credited to meet the shutdown critical safety functions. The inspectors reviewed clearances for outage OR12 and verified, on a sampling basis, that the tagging controls were properly implemented. NRC findings in this area are discussed in this section and section 4OA3 of this report.

Inventory Control The inspectors reviewed FPLE actions to establish, monitor and maintain the proper water inventory in the reactor during the outage, and in the reactor and spent fuel pool after flooding the reactor cavity for refueling activities. The inspectors reviewed the plant system flow paths and configurations established for reactor makeup and verified the configurations were consistent with the outage plan.

Foreign Material Exclusion The inspectors reviewed the implementation of Seabrook procedures for foreign material exclusion control (FME) for the open reactor vessel, reactor cavity and spent fuel pool.

The inspectors reviewed FPLE actions to verify that FME issues were documented and resolved. The inspector interviewed licensee personnel and reviewed condition reports and photographs regarding two foreign material exclusion (FME) issues. One involved foam plugs used in a low pressure turbine extract steam cavity and the other involved the inadvertent introduction of gravel into the B SG. The inspector reviewed FPLE actions to address deficiencies in FME control in the corrective actions system.

Electrical Power The inspectors verified that the status of electrical systems met all TS requirements and FPLEs outage risk control plan. The inspectors verified that compensatory measures were implemented when electrical power supplies were impacted by outage work activities. The inspectors verified that credited backup power supplies were available.

Decay Heat Removal (DHR) System Monitoring The inspectors observed spent fuel pool (SFP) and reactor decay heat removal system status and operating parameters to verify that the cooling systems operated properly.

The review included periodic review of SFP and reactor cavity level, temperature, and RHR flow. The inspectors conducted partial system walkdowns to verify the proper system configuration was established for alternate vessel and cavity level measurement.

Containment Control The inspectors reviewed FPLE activities during the outage to control primary containment closure and integrity, and to prepare the containment for closure prior to plant restart. The inspectors performed walkdowns of all levels in the containment throughout the outage and prior to plant startup per procedure OS1015.18 to review FPLEs cleanup and demobilization controls in areas where work was completed to assure that tools, materials and debris were removed. This review focused on the control of transient combustibles and the removal of debris that could impact the performance of safety systems.

Monitoring Plant Heatup, Approach to Critical and Startup The inspectors observed operator performance during the plant startup activities conducted between April 30 and May 11, 2008. The inspection consisted of control room observations, plant walkdowns and a review of the operator logs, plant computer information, and station procedures. The inspectors observed the approach to critical on May 7, 2008. The inspectors verified, on a sampling basis, that TS, license conditions, and other requirements for mode changes were met. The inspectors verified RCS integrity throughout the restart process by periodically reviewing RCS leakage calculations and by review of systems that monitor conditions inside the containment.

Problem Identification and Resolution The inspectors verified that FPLE was identifying outage related issues and had entered them into the corrective action program. The inspectors reviewed a sample of the corrective actions to verify they were appropriate to resolve the identified issues.

.1 Inadequate Configuration Control - Leak from CS-V-299

Introduction:

A self-revealing non-cited violation of Technical Specification 6.7.1.a was identified for the failure to implement written procedures governing safety-related activities. Specifically, on April 20, 2008, FPLE failed to implement tagging and configuration control procedures. As a result operators established flow through a partially disassembled charging system valve, CS-V-299, resulting in the leak of 200 gallons of reactor cavity water onto the floor of the Primary Auxiliary Building (PAB).

Description:

On April 20, 2008, the reactor was in Mode 6 with the RHR system in service, the reactor open to the reactor cavity, and the cavity flooded to the 14 ft elevation. Maintenance was in progress on CS-V-299 at the time per work order WO 0518410 and Clearance MT005-05A. The valve bonnet and actuator were removed on April 18 to replace the diaphragm. Operators established a charging system lineup per procedures OS 1002.01 and OS 1002.02 to place letdown in service. This lineup placed CS-V-299 in the flow path. As described above CS-V-299 was disassembled. As a result when letdown was placed in service on April 20 reactor water drained out of the valve body and onto the PAB floor. Workers in the immediate area reported the leak to the control room and operators isolated letdown to stop the leak ten minutes after it was initiated.

The leak occurred because of inadequate communication between work groups.

Specifically, the clearance order, which should have prevented operations from placing CS-V-299 in service, was revised to exclude CS-V-299. This change was authorized by the work supervisor because he believed that CS-V-299 was intact; even though he had not verified the actual status of CS-V-299 with the worker performing the maintenance.

The inspectors determined that this was a performance deficiency because the Seabrook clearance tagging administrative procedure, MA 4.2, Step 4.8.2, specified that, in order to revise a clearance tagging boundary, workers performing the work associated with the applicable clearance tagging boundary must be consulted to identify components that must be included in a revised clearance tagging boundary. Contrary to these requirements, on April 20, 2008, the clearance tagging boundary for the CS-V-299 work was revised without consulting the worker performing the work, and, as a result, the integrity of CS-V-299 was not verified before placing it in service.

Analysis:

This finding was more than minor because it was associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of plant events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.

Specifically, the loss of configuration control in the charging system drained 200 gallons from the reactor cavity, which affected the shutdown critical safety function of maintaining adequate reactor inventory, and caused an uncontrolled leak of radioactively contaminated water to a work area.

The finding was determined to be of very low safety significance using the SDP Phase 1 assessment. This issue was evaluated with the assistance of the NRC Region I Senior Reactor Analyst (SRA) using Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process (SDP). The SRA estimated the increase in conditional core damage probability for this event at low E-08. This estimate was derived using IMC 0609, Appendix G, Attachment 2, Significance Determination Process Template for PWR during Shutdown, and considered the following assumptions: 1) The reactor had been shutdown in Mode 6 with the Reactor Cavity partially flooded (14 ft elevation) and the time to boiling was greater than two hours, 2) The leak was very minor and would not have had a significant effect on the volume of water available in either the cavity and/or the RWST. An evaluation of the Appendix G worksheets for plant operating state 2 showed that one sequence was dominant and it involved a loss of inventory with a loss of RCS makeup capability. Both trains of RHR were available throughout the event and would have remained available for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> based on the existing leak rate. The charging and safety injection systems were available during the event and the leak was well within their makeup capability. The issue had very low safety significance (Green) since the finding did not result on a loss of control of shutdown operations and adequate mitigation capability remained available.

The finding has a cross-cutting aspect in the area of human performance, work control, since FPL Energy did not plan and coordinate work activities consistent with nuclear safety (H.3(b)). Specifically, FPLE revised a clearance tagging boundary without verifying the status of affected work activities in accordance with site procedures.

Enforcement:

Technical Specifications 6.7.1.a requires that written procedures be established and implemented to cover the activities described in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Procedure MA4.2 was written pursuant to the above. MA 4.2 requires tagging personnel to contact workers performing work under a clearance to identify components needed to be included in a revised tagging boundary. Contrary to the above, on April 20, 2008, Seabrook did not implement step 4.8.2 of procedure MA4.2. Specifically, clearance MT005 was revised and the letdown system was placed in service without verifying that the physical status of CS-V-299 was appropriate for the flow path or the tagging boundary, resulting in a leak of 200 gallons of water from the reactor cavity to the PAB. Because the finding was of very low safety significance and was entered into the corrective action program as Condition Report 08-06270, this violation is being treated as an NCV, consistent with section VI.A of the NRC Enforcement Policy (05000443/2008003-01, Failure to follow tagging procedure caused inadvertent drain of 200 gallons from RCS).

.2 Inadequate Configuration Control - Leak Through Stow-operated Valve CS-V-1190

Introduction:

A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified because FPLE did not implement corrective actions to prevent recurrence of mispositioned valves caused by difficult to operate stow-operator reach rods. Specifically, on April 20, 2008, a mispositioned (partially open),stow-operated filter drain valve, CS-V-1190, resulted in the inadvertent draining of 2000 gallons of water from the reactor cavity while operators placed the reactor letdown system into service. The drain valve was partially open because it was difficult to operate when positioned with its stow-operator. The mispositioning of a stow-operated valve in a safety system was a repeat occurrence of a similar event in October 2007.

Description:

On April 20, 2008, the reactor was in Mode 6 with the RHR system in service, the reactor open to the reactor cavity, and the cavity flooded to 14 feet above the vessel flange. At about 1:45 p.m., the CVCS was placed in service. Several hours later an operator noticed the sump pump in the PAB basement had pumped approximately 2000 gallons to the floor drain tank. FPLE identified that CVCS filter F-2 drain valve CS-V-1190 was partially open instead of closed as required. The CVCS was shutdown and isolated from the reactor cavity. The operators locally closed valve CS-V-1190 to isolate the leak. The leakage resulted in a loss of approximately 2 inches of level from the reactor cavity.

During a valve lineup to place the CVCS in service, drain valve CS-V-1190 was required to be closed. In fact, the valve was approximately 1.5 turns open, which provided a leak flow of 11 gpm to the PAB floor drain header. The drain valve was operated with a stow-operator reach rod and was difficult to operate. The drain valve was out of position because the Nuclear System Operator who performed the valve lineup believed the valve was shut due to the difficulty operating the valve.

The inspectors determined that the mispositioning of the stow-operated CVCS drain valve was a performance deficiency because it was caused by a condition that should have been corrected by FPLE actions taken in response to a similar past event. In October 2007, a partially open stow-operated drain valve in the RHR system had resulted in continued plant operation with a flow path that bypassed the primary containment boundary (reference Condition Report 200701399). The Seabrook Operating Experience Manual and corrective action program implementing procedure OE3.6 state that deficiencies that could have an effect on plant safety or breach the containment boundary are significant conditions adverse to quality. 10 CFR 50 Appendix B, Criterion XVI requires that corrective actions be taken to prevent recurrence of significant conditions adverse to quality. The corrective actions implemented for the October 2007 containment bypass event did not prevent the reactor cavity drain down event in April 2008. This was a violation of 10 CFR 50 Appendix B, Criterion XVI.

Analysis:

This finding was more than minor because it was associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of plant events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.

Specifically, the loss of configuration control in the charging system unintentionally drained 2000 gallons from the reactor cavity, which affected the shutdown critical safety function of maintaining adequate reactor inventory.

The finding was determined to be of very low safety significance using the SDP Phase 1 assessment. This issue was evaluated with the assistance of the NRC Region I Senior Reactor Analyst (SRA) using Inspection Manual Chapter (IMC) 0609, Appendix G, Shutdown Operations Significance Determination Process (SDP). The SRA estimated the increase in conditional core damage probability for this event at low E-08. This estimate was derived using IMC 0609, Appendix G, Attachment 2, Significance Determination Process Template for PWR during Shutdown, and considered the following assumptions: 1) The reactor had been shutdown in Mode 6 with the Reactor Cavity partially flooded (14 ft elevation) and the time to boiling was greater than two hours, 2) The leak would not have had a significant effect on the volume of water available in either the cavity and/or the RWST. An evaluation of the Appendix G worksheets for plant operating state 2 showed that one sequence was dominant and it involved a loss of inventory with a loss of RCS makeup capability. Both trains of RHR were available throughout the event and would have remained available for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> based on the existing leak rate. The charging and safety injection systems were available during the event and the leak was well within their makeup capability. The issue had a very low safety significance (Green) since the finding did not result on a loss of control of shutdown operations and adequate mitigation capability remained available.

The finding has a cross-cutting aspect in the area of problem identification and resolution because FPL Energy did not take appropriate corrective actions to address safety issues in a timely manner commensurate with their safety significance and complexity (P.1.d).

Specifically FPL Energy did not take adequate corrective actions to assure the correct positioning of stow-operated safety system valves and thereby prevent recurrence of a significant condition adverse to quality.

Enforcement:

10 CFR 50 Appendix B, Criterion XVI requires that corrective actions be taken to prevent recurrence of significant condition adverse to quality. Contrary to the above, FPLE did not implement corrective actions to prevent recurrence of mispositioned valves caused by difficult to operate stow-operator reach rods.

Specifically, on April 20, 2008, a mispositioned (partially open), stow-operated filter drain valve, CS-V-1190, resulted in the inadvertent draining of 2000 gallons of water from the reactor cavity while operators placed the reactor letdown system into service. The drain valve was partially open because it was difficult to operate when positioned with its stow-operator. The mispositioning of a stow-operated valve in a safety system was a repeat occurrence of a similar event in October 2007. Because the finding was of very low safety significance and has been entered into the corrective action program as Condition Report 08-06260, this violation is being treated as a NCV, consistent with section VI.A of the NRC Enforcement Policy (NCV 05000443/2008003-02, Inadequate corrective actions to prevent recurrence of mispositioned stow-operated valves caused inadvertent drain of 2000 gallons from RCS).

1R22 Surveillance Testing

a. Inspection Scope

The inspectors completed six surveillance testing inspection samples. The inspectors observed portions of surveillance testing activities of safety-related systems to verify that the system and components were capable of performing their intended safety function, to verify operational readiness, and to ensure compliance with required TS and surveillance procedures.

The inspectors attended selected pre-evolution briefings, performed system and control room walkdowns, observed operators and technicians perform the test evolutions, reviewed system parameters, and interviewed the applicable system engineers and field operators. The test data recorded was compared to procedural and technical TS requirements, and to prior tests results to identify any potential adverse trends. The following surveillance procedures were reviewed.

• Centrifugal Charging Pump Comprehensive Pump Test per procedure OX1456.92 performed on April 2, 2008

• LLRT of Penetration X-38B (Combustible Gas Control) performed per WO 0700117 on April 2, 2008

• LLRT of Penetration X-72C (Combustible Gas control Purge Return) performed per WO 0700163 on April 3, 2008

• Reactor Containment Integrated Leakage Rate Test - Type A performed per procedure EX1803.001 between April 25 and April 28, 2008

• Diesel Generator 1A/1B 18 Month Operability and Engineered Safeguards Pump and Valve Response Time Testing Surveillance performed per procedure OX1426.19/20 between April 30 and May 5, 2008

• Emergency Diesel Generator 1A 24 Hour Load Test and Hot Restart Surveillance, OX14.26.22, Rev. 01, Chg. 06 performed between June 4 and June 5, 2008.

The inspectors reviewed deficiencies related to surveillance testing and verified that the issues were entered into the corrective action program. The references used for this review are listed in Attachment A.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access to Radiological Significant Areas (71121.01 - 10 samples)

a. Inspection Scope

During the period April 14 and17, 2008, the inspectors conducted the following activities to verify that FPLE was properly implementing physical, administrative, and engineering controls for access to locked high radiation areas, and other radiological controlled areas (RCA) during the refueling outage (OR12). The inspectors also verified that workers were adhering to these controls when working in these areas. Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, Seabrook Technical Specifications, and Seabrook=s procedures.

This activity represents the completion of ten samples for this inspection area.

Plant Walkdown and RWP Reviews The inspectors identified exposure significant work areas in the Containment Building and Primary Auxiliary Building (PAB) for ongoing outage activities. Tasks in the Containment Building included using an advanced scale conditioning agent (ASCA) for cleaning the secondary side of the steam generators, performing weld overlays on pressurizer nozzles, emergency core cooling system sump modifications, preparations for cavity decontamination, and various support work including scaffolding erection and insulation removal. Tasks in the PAB included inspection and maintenance of the B residual heat removal (RHR) system. The inspectors reviewed the radiation work permits (RWP) and the radiation survey maps associated with these work areas to determine if the radiological controls were acceptable.

The inspectors toured accessible radiological controlled areas located in the Containment Building, Primary Auxiliary Building, Decay Heat Vaults, Fuel Storage Building, and Waste Processing Building, with radiation protection supervision. The inspector performed independent radiation surveys in these areas to confirm the accuracy of survey maps and the adequacy of postings and barricades.

In reviewing RWPs, the inspectors evaluated electronic dosimeter (ED) locations on personnel and dose/dose rate alarm setpoints to determine if ED placement was in the highest dose field and that the setpoints were consistent with the area radiological conditions and plant policy. The inspectors verified that workers were knowledgeable of the actions to be taken when the electronic dosimeter alarms or malfunctions. Work activities reviewed included, scaffold erection (RWP 08-038), steam generator ASCA operations (RWP 08-033), cavity decontamination (RWP 08-028) and valve maintenance (RWP 08-036).

The inspectors reviewed the radiological controls applied to recently completed outage tasks to evaluate the effectiveness of controlling exposure. Included in the review were regenerative heat exchanger maintenance (RWP 08-050), reactor head lift (RWP 08-001), and removal of the D reactor coolant pump motor (RWP 08-034).

Problem Identification and Resolution The inspectors reviewed elements of Seabrook=s corrective action program related to controlling access to radiological controlled areas and completed since the last inspection of this area to determine if problems were entered into the program for resolution. The inspectors reviewed daily quality summaries, a radiation control program audit (SBK-08-01), condition reports, and associated apparent cause evaluations.

Additionally, the inspectors reviewed dose and dose rate alarm reports and dosimetry abnormality occurrence reports to verify that no performance indicator or regulatory limit was exceeded.

Jobs-In-Progress The inspectors observed aspects of various outage related tasks performed during this inspection period to verify that radiological controls, such as required surveys, area postings, job coverage, and pre-job RWP briefings were appropriately conducted; personnel dosimetry was appropriately worn; and that workers were knowledgeable of work area radiological conditions. Tasks observed included preparations for reactor cavity decontamination, containment sump modifications, and pressurizer weld overlays.

High Risk Significant, High Dose Rate HRA, and VHRA Controls The inspectors discussed with the Radiation Protection Manager and senior technicians high radiation area (HRA) and very high radiation area (VHRA) controls and procedures.

These special areas included under reactor vessel areas and spent fuel transfer routes in containment, spent resin sluicing paths and spent resin storage locations in the PAB, and irradiated hardware stored in the spent fuel pool. The inspectors evaluated the pre-requisite communications, procedural authorizations, and operational controls that must be implemented prior to conducting activities in these plant areas. The inspectors verified that any changes to relevant procedures did not substantially reduce the effectiveness and level of worker protection.

Keys to locked high radiation areas (LHRA) and VHRAs, maintained at the radiation protection control point and in the alternate control point, were inventoried, and accessible LHRAs were verified to be properly secured and posted during plant tours.

Radiation Worker/Radiation Protection Technician Performance The inspectors observed radiation worker and radiation protection technician performance by attending various pre-job/RWP briefings, observing activities in progress, and questioning individuals regarding their knowledge of radiological controls and contamination control measures that applied to their tasks when working in the RCA.

The inspectors reviewed conditions reports related to radiation worker and radiation protection technician errors to determine if an observable pattern traceable to a common cause was evident.

b. Findings

Introduction:

A Green, self-revealing NCV of Technical Specification 6.11.2 was identified. Specifically, on May 1, 2008, FPLE did not identify and control an existing high radiation area with dose rates greater than 1000 millirems per hour in the reactor containment building. A worker was exposed to higher than expected radiation levels of approximately 2,270 mrems per hour.

Description:

On May 1, 2008, a worker entered the reactor containment building to adjust damper CAH-638, located in the pressurizer surge line chase area. The workers electronic alarming dosimeter unexpectedly alarmed when he was exposed to unanticipated radiation levels of approximately 2,270 millirem per hour. Subsequent surveys at the source of radiation around the pressurizer surge line measured 10,000 millirem per hour on contact and 4,000 millirem per hour at 30 centimeters. The area was not barricaded, conspicuously posted, or guarded as a locked high radiation area.

Upon receiving the dose rate alarm (setpoint of 500 millirem per hour), the worker immediately left the area. FPLE determined that the worker received a dose of 4 millirem.

FPLE completed flushes earlier in OR12 to reduce the dose rates in the bottom of the pressurizer in preparation for outage work activities. The flushes resulted in higher dose rates in the horizontal section of the pressurizer surge line that was posted and controlled as a locked high radiation area earlier in the outage. Dose rates declined during flood up of the reactor coolant system on April 23 and the area was controlled consistent with a high radiation area based on surveys taken from April 23-29, 2008.

During the preparations for plant startup, FPLE completed operating activities to fill and vent the RCS, bump the reactor coolant pumps (RCPs), and operate the C RCP for 30 minutes on May 1, 2008. The operating activities had the potential to relocate the radiological source term in the pressurizer surge line. Although FPLE surveyed the RCS to monitor changes in dose rates caused by the restart activities, FPLE did not survey the pressurizer surge line. FPLE identified that the surge line radiation levels had increased and re-established locked high radiation area controls based on surveys taken after the worker received an unexpected electronic dosimeter alarm upon entering the area.

Analysis:

The failure to control access to a high radiation area is a performance deficiency. The finding is more than minor because it is associated with the occupational radiation safety cornerstone attribute of exposure control and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, not controlling a locked high radiation area in accordance with TS requirements could increase personnel exposure. Because this occurrence involved an unintended dose or potential for dose that could have been significantly greater as a result of a single minor, reasonable alteration of circumstances, the significance of this finding was evaluated using the occupational radiation safety significant determination process. The inspectors determined that the finding was of very low safety significance (Green) because it did not involve ALARA planning and controls, an overexposure, a substantial potential for overexposure, or an impaired ability to assess dose. This finding was entered into FPLEs corrective action program as Condition Report CR200806982.

This finding had a cross-cutting aspect in the area of human performance, work control, because FPLE did not appropriately plan work by incorporating job site conditions that impact radiological safety. Specifically, FPLE did not adequately assess changing area dose rates caused by operating activities, and thus did not adequately plan a work task with due consideration of the actual radiological conditions at the job site (H.3(a)).

Enforcement:

Technical Specification 6.11.2, states, in part, that for individual high radiation areas with radiation levels greater than or equal to 1000 millirem per hour that are accessible to personnel, that are located within large areas such as a reactor containment, where no enclosure exists for purposes of locking, or that is not continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device. Contrary to the above, on May 1, 2008, FPLE did not properly identify and control a high radiation area with dose rates greater than 1000 millirem per hour. Specifically, FPLE did not adequately assess changing area dose rates in the pressurizer surge line chase area that were caused by operating activities, and thus did not identify that area as a high radiation area with radiation levels greater than or equal to 1000 millirem per hour and therefore did not implement the required radiological controls for that area. Because the failure to control a high radiation area as a locked high radiation area was determined to be of low safety significance (Green), and was entered into FPLE's corrective action program as CR 08-06982, this violation is being treated as an NCV consistent with Section VI.A of the NRC Enforcement Policy, NUREG-1600. (NCV 05000443/2008003-03, failure to control a high radiation area as a locked high radiation area)2OS2 ALARA Planning and Controls (71121.02 - 17 samples)

a. Inspection Scope

During the period April 14 to 17, 2008, the inspectors conducted the following activities to verify that FPLE was properly implementing operational, engineering, and administrative controls to maintain personnel exposure as low as is reasonably achievable (ALARA) for tasks conducted during the refueling outage (OR12).

Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, applicable industry standards, and FPLE=s procedures.

This inspection activity represents completion of seventeen samples for this inspection area.

Radiological Work Planning The inspectors reviewed pertinent information regarding the site=s cumulative exposure history, current exposure trends, and ongoing activities to assess current performance and exposure challenges. The inspectors determined the plant=s three-year rolling collective average exposure and concluded that the site was ranked in the top performance quartile for U.S. pressurized water reactors.

The inspectors reviewed the refueling outage work scheduled during the inspection period and the associated work activity exposure estimates. Scheduled work included steam generator secondary side cleaning, reactor cavity decontamination, pressurizer weld overlays, containment sump modifications, and valve maintenance. As part of this review, the inspectors evaluated the dose estimates for these jobs and reviewed the associated ALARA Plans. The inspectors also reviewed the procedures associated with maintaining worker dose ALARA and with estimating and tracking work activity specific exposures.

The inspectors reviewed the daily OR12 Project Dose Summary Report, detailing the worker estimated and actual exposures, through April 17, 2008, for jobs performed during the refueling outage.

The inspectors evaluated the exposure mitigation requirements, specified in ALARA Reviews (AR), and compared actual worker cumulative exposure to estimated dose for tasks associated with these work activities. Jobs reviewed included reactor vessel dis-assembly/re-assembly (AR 08-01), steam generator secondary side maintenance (AR 08-02), in-service inspection (AR 08-03), cavity decon (AR 08-04), valve maintenance (AR 08-06), scaffolding installation/removal (AR 08-10), pressurizer weld overlay project (AR 08-12), and containment sump modification (AR 08-15).

The inspectors evaluated the departmental interfaces between radiation protection, operations, maintenance crafts, and engineering to identify missing ALARA program elements and interface problems. The evaluation was accomplished by interviewing the Radiation Protection Manager and the ALARA Coordinator, reviewing Radiation Safety Committee meeting minutes, reviewing outage-related Nuclear Assurance Daily Quality Summary Reports, observing jobs-in-progress, and attending the pre-job briefing for reactor cavity decontamination.

The inspectors compared the person-hour estimates provided by the maintenance planning and other work groups with actual work activity time requirements and evaluated the accuracy of these time estimates. Specific work activities evaluated included pressurizer weld overlay, scaffolding installation, containment sump modification, and steam generator secondary side cleaning.

The inspectors determined if work activity planning included the use of remote audio/video monitoring, temporary shielding, system flushes, relocation of irradiated components away from occupied work areas, and operational considerations to further minimize worker dose. In doing this evaluation, the inspector reviewed temporary shielding requests, cavity decontamination pre-requisites, shutdown chemistry requirements, and steam generator preparations for cleaning.

Verification of Dose Estimates and Exposure Tracking Systems The inspectors reviewed the assumptions and basis for the current annual collective exposure estimates for the operating cycle and refueling outage and compared this to actual exposure data.

The inspectors reviewed FPLE=s method for adjusting exposure estimates, and re-planning work, based on work progress. This review included evaluating the basis for the Radiation Safety Committee establishing the outage stretch goal of 72 person-rem compared to a business plan goal of 78 person-rem.

The inspectors reviewed FPLE=s exposure tracking system to determine whether the level of dose tracking detail, exposure report timeliness, and exposure report distribution was sufficient to support the control of collective and individual exposures. Included in the review were electronic dose and dose rate alarm reports, departmental collective exposure data, and identification of the highest individual dose receptors.

Job Site Inspection and ALARA Control The inspectors observed maintenance and operational activities performed for steam generator secondary side cleaning, reactor cavity decontamination, containment building demobilization, and pressurizer weld overlay to verify that pre-requisite radiological controls were implemented and workers were knowledgeable of work area radiological conditions and ALARA practices.

The inspectors reviewed the exposures for selected individuals in various work groups, including electrical maintenance, radiation protection, contractors, and mechanical maintenance to determine if supervisory efforts were made to equalize dose among the workers.

Source Term Reduction Control The inspectors reviewed the current status and historical trends of the site=s source terms. Through interviews with the Chemistry Supervisor and Radiation Protection Manager, the inspectors evaluated the effectiveness of FPLE=s source term control strategy. Specific strategies employed by FPLE included post-shutdown peroxide flushes of the reactor coolant system, use of a macroporous resin for coolant cleanup, use of a submersible demineralizer for reactor cavity cleanup, relocating irradiated components away from work areas, and customized temporary shielding for the pressurizer surge line Radiation Worker Performance The inspectors observed radiation worker and health physics technician performance during pressurizer weld overlays at the centralized monitoring station. The inspectors determined whether the individuals were aware of current radiological conditions, access controls, and that the skill level was sufficient with respect to effectively performing their tasks and implementing proper ALARA practices.

The inspectors attended the pre-job briefing for a exposure significant task, reactor cavity decontamination. The inspectors determined that roles and responsibilities were identified, that the sequencing of various activities were iterated, and that lessons learned from past cavity decontamination tasks were reviewed.

The inspectors reviewed condition reports, related to radiation worker and radiation protection technician errors, and personnel contamination reports (PCR) to determine if an observable pattern traceable to a similar cause was evident.

Declared Pregnant Workers The inspectors determined that there were no declared pregnant workers performing outage related activities in the RCA during the inspection period.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification (71151 - 1 Sample)

.1 Safety System Functional Failures

a. Inspection Scope

The inspectors sampled FPLE submittals for the performance indicators (PIs) listed below for the period from January 2007 through December 2007. To verify the accuracy of the PI data reported during that period, PI definitions and guidance contained in NEI 99-02, "Regulatory Assessment Indicator Guideline," Rev. 5 were used to verify the basis in reporting for each data element.

Mitigating Systems Cornerstone

• Safety system Functional Failures The inspectors reviewed plant records such as Licensee Event Reports (LERs),operating logs, procedures, and interviewed applicable licensee personnel to verify the accuracy and completeness of Seabrook's PI data. The inspectors also reviewed the accuracy of the number of critical hours reported.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152 - 1 Samples)

.1 Routine Condition Report Screening

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the Seabrooks corrective action program. This review was accomplished by accessing Seabrook's computerized database.

b. Findings

No findings of significance were identified.

.2 Semi-annual Review to Identify Trends

a. Inspection Scope

As required by Inspection Procedure 71152, Problem Identification and Resolution, the inspectors performed a review of Seabrooks CAP and associated documents to identify trends that may indicate existence of safety significant issues. The inspectors review was focused on repetitive equipment and corrective maintenance issues, but also considered the results of daily CAP item screening. The inspectors compared and contrasted their results with the results contained in the Seabrook CAP Quarterly Trend Reports.

b. Findings

No findings of significance were identified. The inspectors did not identify any appreciable trends that Seabrook had not already identified.

4OA3 Event Follow Up

a. Inspection Scope

The inspectors completed four event follow-up inspection samples. The inspector reviewed FPLE actions during the unplanned, non-routine events listed below. The inspection focused on personnel performance and was accomplished through interviews of personnel, observation of the work site and fact-finding meetings, and reviews of operator logs, plant computer information, alarm printouts, and station drawings and procedures. The inspector reviewed the adequacy of personnel performance, the use of abnormal and emergency procedures, and corrective actions. Operator actions were compared to station procedures to determine if the response was appropriate and in accordance with procedures and training. The plant equipment response was reviewed and verified acceptable or captured in the corrective action program. Documents reviewed during this inspection are listed in Attachment A. The events included:

• The response of the control room staff and the fire brigade per OS 1200.00 to a small fire in the pressurizer cubicle on April 3, 2008. The assigned fire watch extinguished the fire using a portable fire extinguisher. FPLE evacuated the containment pending a review of the work site, the event causes, and the extent of condition.

• The response to a loss of water from the reactor cavity while placing the letdown system in service at 4:22 am on April 20, 2008. The leak was through partially disassembled charging system valve CS-V-299. This item is discussed further in Section 1R20.b.1 of this report.

• The response to a loss of water from the reactor cavity while placing the letdown system in service at 5:00 pm on April 20, 2008. The leak was through partially open charging system drain valve CS-V-1190. This item is discussed further in Section

1R20 .b.2 of this report.

• The plant load decrease per OS1230.04 to 30% full power in response to a condenser tube leak on June 5, 2008. The C condenser waterbox was isolated per ON1038.08 pending identification and repair of a single tube in the north water box.

The plant was returned to full power on June 8 upon completion of repairs and restoration of chemistry parameters within acceptable limits.

b. Findings

No findings of significance were identified.

4OA5 Other Activities

.1 Inspection Results for TI 2515/172, RCS Dissimilar Metal Butt Welds

a. Inspection Scope

Temporary Instruction, TI 2515/172, provides for confirmation that owners of pressurized-water reactors (PWRs) have implemented the industry guidelines of the Materials Reliability Program (MRP)-139 regarding nondestructive examination and evaluation of certain dissimilar metal welds in reactor coolant systems containing Alloy 600/82/182. The TI requires documentation of specific questions in this inspection report. The questions and responses are included in Attachment B to this report.

In summary, the Seabrook Station pressurizer has six dissimilar metal welds (one 14 surge line nozzle, one 4 spray nozzle and four 6 safety/relief nozzles). Also, there are four RCS hot leg (HL) outlet nozzles and four RCS cold leg (CL) inlet nozzles on the reactor vessel (RV) which are MRP-139 applicable Alloy 600/82/182. Seabrook Station has submitted an Alternative Request that is applicable to these welds (excluding the eight RV inlet and outlet nozzles) to allow the performance of a preemptive full structural weld overlay on the pressurizer surge, spray, and safety line welds. The proposed alternative (SBK-L-07120, dated 07/03/2007) and supplement (SBK-L-08022, dated 12/13/2008) were approved on April 1, 2008, by NRC Staff.

b. Findings

No findings of significance were identified.

.2 Temporary Instruction 2515/166 - Pressurized Water Reactor Containment Sump

Blockage (NRC Generic Letter 2004-02)

a. Inspection Scope

The inspectors performed an inspection in accordance with Temporary Instruction (TI)2515/166, Pressurized Water Reactor Containment Sump Blockage, Revision 1. The TI was developed to support the NRC review of licensee activities in response to NRC Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors. Specifically, the inspectors verified that the implementation of the modifications and procedure changes was consistent with the actions committed to in FPLEs responses to GL 2004-02.

The inspectors reviewed a sample of the licensing and design documents to verify that they were updated, or in the process of being updated, to reflect the modifications to the plant. The inspectors performed field walkdowns of the strainer installation to verify that it was performed in accordance with the approved design change package, and to verify FPLEs conclusion of no containment choke-points that could prevent water from reaching the recirculation sump during a design basis accident. The inspectors discussed details of the containment sump modification with engineers, project managers, and field installation supervisors to verify design control of the modification process. Finally, the inspectors reviewed FPLE procedures for final acceptance and foreign material inspection of the sump, as well as procedures for containment coatings inspections and final containment closeout, to evaluate adequacy. Documents reviewed are listed in the Attachment A.

b. Evaluation of Inspection Requirements The TI required the inspectors to evaluate and answer the following questions:

Did the licensee implement the plant modifications and procedure changes committed to in their GL 2004-02 response?

The inspectors verified that FPLE implemented the plant modifications and procedure changes committed to in their GL 2004-02 responses. The inspectors verified installation of the containment sump strainer and verified the strainer surface area was consistent with the GL response. The inspectors verified that the installed modification met the assumptions of FPLEs testing and analyses, including chemical effects and downstream effects. Finally, the inspectors reviewed various procedure changes to verify that the assumptions described in FPLEs GL responses were valid.

Has the licensee updated its licensing basis to reflect the corrective actions taken in response to GL 2004-02?

The inspectors verified that changes to the facility and procedures as described in the Updated Final Safety Analysis Report (USFAR), and identified in FPLEs GL 2004-002 responses, were reviewed and documented in accordance with 10 CFR 50.59. Additionally, the inspectors verified that FPLE had either updated, or was in the process of updating, the licensing basis to reflect the actions taken in response to GL 2004-02. Specifically, the required changes to the UFSAR were in the process of being updated at the time of inspection. No license amendments were required.

The inspection requirements of the TI are complete and the TI is closed. FPLE is committed to a final supplemental response within 90 days after completion of refueling outage OR12 (Spring 2008). The response will provide the remaining information regarding issues discussed in the GL, including results of FPLEs recently completed downstream effects evaluations, as well as chemical effects testing and analysis.

This documentation of TI-2515/166 completion as well as any results of sampling audits of licensee actions will be reviewed by the NRC staff (Office of Nuclear Reactor Regulation - NRR) as input along with the Generic Letter (GL) 2004-02 Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors responses to support closure of GL 2004-02 and Generic Safety Issue (GSI)-191 Assessment of Debris Accumulation on Pressurized-Water Reactor (PWR) Sump Performance." The NRC will notify FPLE by letter of the results of the overall assessment as to whether GSI-191 and GL 2004-02 have been satisfactorily addressed at Seabrook Station.

.3 (Closed) URI 2008002-01: Inaccurate Information in Initial Operator License Application

During the previous reporting period, the NRC issued an unresolved item to document a concern regarding FPLEs notification to the NRC of the identification on January 28, 2008, of inaccurate information provided on an application for a senior reactor operator (SRO) license. The issue was described in Section 4OA5 of NRC Inspection Report 05000443/2008002. During this inspection period, the inspector reviewed FPLE actions as described in Section 4OA7 of this report. URI 2008002-01 is closed.

4OA6 Meetings, including Exit

The inspectors presented the inspection results to Mr. Gene St. Pierre on July 1, 2008, following the conclusion of the period. FPLE did not indicate that any of the information presented at the exit meeting was proprietary.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Severity Level IV) was identified by FPLE and is a violation of NRC requirements that meets the criteria of Section VI of the NRC Enforcement Policy, for being dispositioned as an non-cited violation (NCV).

• 10 CFR 50.9 requires, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on June 8, 2007, FPLE submitted a NRC Form 398 application for an individual=s senior reactor operator license that was not complete and accurate in all material respects. Specifically, the application indicated the individual met the requirement for three years of responsible power plant experience; however, this was inaccurate because the individual had less than the three years of responsible power plant experience. This information was material to the NRC because the NRC used the information submitted on the 368 to allow the applicant to take the initial license exam, and ultimately, issue the individual an SRO license. The traditional enforcement process was used to disposition the violation because it impacted the NRC=s ability to perform its regulatory function.

The finding was more than minor because it was a non-willful compromise of an application required by 10 CFR Part 55 that contributed to an individual being granted a SRO license. The violation was licensee identified via an internal audit and entered into their corrective action program (CR 08-01388). FPLE performed a root cause evaluation and informed the NRC. The finding was of very low safety significance because the licensed individual properly performed licensed duties and because the NRC would most likely have granted a waiver of experience requirements, based on the applicant=s work history, had a waiver been requested. (05000443/200800304, Inaccurate Information on Initial Operator License Application, EA-08-164).

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

P. Allen, Senior Health Physics Technician

M. Ames. Nuclear Plant Operator

R. Arns, Engineering

J. Ball, Maintenance Rule Coordinator

R. Belanger, Design Engineer

D. Berko, Director Plant Support

M. Bianco, Supervisor, Radiological Waste Services

B. Brown, Plant Engineer

K. Browne, Assistant Operations Manager

J . Burson, Operations Training Instructor

L. Carlsen, Operations Training Instructor

W. Cash, Chemistry Manager

D. Chang, Tagging Support

B. Clark, Radiation Protection Supervisor

R. Couture, Reactor Engineer

J. Crowley, I&C Superintendent

T. Date, Senior Health Physics Technician

J. Desmond, Operations Training Instructor

D. Egonis, Plant Engineer

D. Feeney, Mechanical Maintenance

D. Flahardy, Radiation Protection Supervisor

P. Freeman, Engineering Director

D. Hampton, Health Physics Shift Supervisor

M. Haskins, Maintenance Manager

S. Kessinger, OC Suprervisor

G. Kim, Risk Analyst

E. Metcalf, Operations Manager

J. Kennish, Operations Training Instructor

M. Kiley, Plant Manager

M. Lipman, Plant Technician

T. Manning, Engineering

D. Master, Plant Engineer

D. Masters, Engineering

B. McAllister, SW System Engineer

N. McCafferty, Plant Engineering Manager

E. Metcalf, Operations Manager

M. OKeefe, Regulatory Compliance Supervisor

K. Mahoney, Reactor Engineer

E. Metcalf, Operations Manager

R. Noble, Engineering Manager

J. Peschel, Regulatory Programs Manager

E. Piggot, Unit Supervisor

R. Plante, Maintenance Supervisor.

N. Pond, Tagging Coordinator

K. Purington, Reactor Operator

K. Randall, Reactor Engineer

T. Rossengal, RHR System Engineer

M. Russell, Operations Clerk

M. Scannell, Health Physics Shift Supervisor - Nuclear

W. Schoppmeyer, Nuclear Oversight Assessor

D. Sherwin, Maintenance Assistant

D. Skiffington, Containment Sump Field Installation Supervisor

J. Soucie, NPO

E. Spader, Training Supervisor

R. Sterritt, Health Physics Specialist - Nuclear

M. Taylor, Shift Manager

K. Thibodeau, Operations Training Instructor

R. Thurlow, Radiation Protection Manager

J. Tucker, Security Manager

J. Varga, Reactor Operator

J. Walsh, CVCS System Engineer

N. Walts, Unit Supervisor

S. Wellhofer, Site Nurse

B. White, Project Engineering Manager

R. White, Security Supervisor

K. Wright, Training Manager

B. Plummer, Nuclear Projects

W. Schmidt, Electrical Maintenance

G. St. Pierre, Site Vice President

J. Varga, Reactor Operator

NRC Personnel

T. Moslak, Health Physicist

• Attended the Exit Meeting on April 17, 2008

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Opened and Closed

05000443/200800301 NCV Failure to follow tagging procedure caused inadvertent drain of 200 gallons from RCS.. (Section 1R20b.1)

05000443/200800302 NCV Inadequate corrective actions to prevent recurrence of mispositioned stow-operated valves caused inadvertent drain of 2000 gallons from RCS.. (Section 1R20b.2)

05000443/200800303 NCV Failure to control a high radiation area as a locked high radiation area. (Section 2OS1)

Closed

05000443/200800201 URI Inaccurate Information on Initial Operator License Application (Section 4OA5)