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

Text

August 1, 2008

EA-08-164 Mr. Gene S Site Vice President FPL Energy Seabrook, LLC Seabrook Station c/o Mr. Michael O'Keefe P.O. Box 300 Seabrook, NH 03874

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 Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents 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 2 NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/ 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 cc w/encl: J. A. Stall, FPL Senior Vice President, Nuclear & CNO M. Warner, Vice President, Nuclear Operations R. S. Kundalkar, FPL Vice President, Nuclear Technical Svcs M. Mashhadi, Senior Attorney, Florida Power & Light Company M. S. Ross, Managing Attorney, Florida Power & Light Company M. O'Keefe, Manager, Regulatory Programs P. Freeman, Plant General Manager K. Wright, Manager, Nuclear Training, Seabrook Station R. Poole, FEMA, Region I Office of the Attorney General, Commonwealth of Mass K. Ayotte, Attorney General, State of NH O. Fitch, Deputy Attorney General, State of NH P. Brann, Assistant Attorney General, State of Maine R. Walker, Director, Radiation Control Program, Dept. of Public Health, Commonwealth of MA C. Pope, Director, Homeland Security & Emergency Management, State of NH J. Giarrusso, MEMA, Commonwealth of Mass D. O'Dowd, Administrator, Radiological Health Section, DPHS, DHHS, State of NH J. Roy, Director of Operations, Massachusetts Municipal Wholesale Electric Company T. Crimmins, Polestar Applied Technology R. Backus, Esquire, Backus, Meyer and Solomon, NH Town of Exeter, State of New Hampshire Board of Selectmen, Town of Amesbury S. Comley, Executive Director, We the People of the United States R. Shadis, New England Coalition Staff M. Metcalf, Seacoast Anti-Pollution League

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

Sincerely,/RA/

Arthur L. Burritt, Chief Projects Branch 3 Division of Reactor Projects Distribution w/encl:

(via e-mail)

S. Collins, RA M. Dapas, DRA D. Lew, DRP J. Clifford, DRP A. Burritt, DRP L. Cline, DRP W. Raymond, DRP, Sr. Resident Inspector J. Johnson, DRP Resident Inspector S. Williams, RI OEDO H. Chernoff, NRR R. Nelson, NRR E. Miller, NRR, PM R. Ennis, NRR, Backup T. Valentine, NRR ROPreports@nrc.gov Region I Docket Room (with concurrences)

SUNSI Review Complete: ____ALB______(Reviewer's Initials)

DOCUMENT NAME: G:\DRP\BRANCH3\INSPECTION\REPORTS\ISSUED\SEA0803.DOC After declaring this document "An Official Agency Record" it will be released to the Public.

To receive a copy of this document, indicate in the box:

" C" = Copy without attachment/enclosure " E" = Copy with attachment/enclosure " N" = No copy ML082140855 OFFICE RI/DRP RI/DRP RI/ ORA RI/DRP NAME WRaymond/ LCline/ RSummers/ ABurritt/ DATE 07/14/08 08/01/08 08/01/08 08/01/08 OFFICIAL RECORD COPY U. S. NUCLEAR REGULATORY COMMISSION 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

2 Enclosure

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).

B. 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 FPLE's 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 FPLE's 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 Attachment 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 FPLE's 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 Attachment 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 system's 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 Seabrook's 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 Seabrook's 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 FPLE's 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 FPLE's boric acid control program. The inspectors selected CR's that identified both active and inactive leak locations that could result in degradation of safety significant components. The inspectors reviewed six CR's 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 FPLE's 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 FPLE's conclusion that no SG tube inspection was required for this outage. The inspector noted FPLE's technical evaluation and determination that there were no degradation mechanisms in the Seabrook SG's 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 operator's 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 simulator's 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 evaluator's 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 FPLE's 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 licensee's 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 FPLE's 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:

• WO's 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

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

• WO's 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 Seabrook's 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 FPLE's 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 FPLE's 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 FPLE's 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 FPLE's 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 FPLE's 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 worker's 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 FPLE's 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 Seabrook's 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 Seabrook's 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 FPLE's 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 FPLE's 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:

1. 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 FPLE's testing and analyses, including chemical effects and downstream effects. Finally, the inspectors reviewed various procedure changes to verify that the assumptions described in FPLE's GL responses were valid.

2. 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 FPLE's 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 FPLE's 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 FPLE's 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. O'Keefe, Regulatory Compliance Supervisor

K. Mahoney, Reactor Engineer

E. Metcalf, Operations Manager

R. Noble, Engineering Manager

Attachment

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

Attachment

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/FIN-2008002-01 URI Inaccurate Information on Initial Operator License Application (Section 4OA5)

Section 1R01: Adverse Weather Protection

UFSAR Section 9.4, Air Conditioning, Heating, Cooling And Ventilation Systems

Condition Reports

200807958,

200806913,

Work Order

0727292,

0726471,

0707698,

0722213,

0738946,

0800223,

0807902

ON1490.09, Summer Readiness Surveillance, Revision 1 OS1200.03, Severe Weather Conditions, Revision 13 Work Order Overview Report, 4/29/2008 Alarm Response Procedures for Degraded Power OS1246.02, Degraded vital AC Power (Plant Operating), Revision 5

UFSAR Section 9.2.2, Cooling Systems for Reactor Auxiliaries, Revision 12 Alarm Response Procedure for D6667, 345KV System Trouble Alarm Response Procedure for B8470, 345KV Line 394 Voltage Low Alarm Response Procedure for D6670, 345KV Line Loss of Voltage

Section 1R04: Equipment Alignment

OS1013.04, Residual Heat Removal Train B Startup and Operation, Revision 12 PID 1-RH-B20663, Residual Heat Removal Sys Train B Cross Tie Detail, Revision 18 RHR System Health Report UFSAR Section 5.4.7 Residual Heat Removal System UFSAR Section 6.3, Emergency Core Cooling System Attachment Technical Specifications (TS) Section 3.5, Emergency Core Cooling Systems TS Section 3.4.9.3, Reactor Overpressure Protection

OS1001.11, RCS Shutdown Level Instrumentation, Revision 1

Section 1R05: Fire Protection

Fire Hazards Analysis UFSAR Section 9.5.1 Fire Protection Systems OX443.75, Establishing Containment Fire Protection, Rev. 03 Chg. 08 Fire Protection Pre-fire Strategies Fire Drill Evaluation / Scenario for June 10, 2008 Announced Fire Drill

WO 0732110

Fire Brigade Ready Area Monthly Inventory (prior to drill)

WO 0820947

Fire Brigade Ready Area Monthly Inventory (post drill) ON0443.35 R5 Fire Brigade Ready Area Inventory OS1200.00 R13 Response to Fire or Fire Alarm Actuation ER 1.1 R45

Classification of Emergencies

CR 08-08206

Fire brigade members have expired SCBA training in Sentinel Data base

CR 08-08297

Evaluation of "Tech Spec" extensions for extending training

CR 08-08205

During a fire drill wrong size respirator worn by a fire brigade member

CR 08-08576

Recently qualified Fire Brigade Leader required augmented shift coverage

Section 1R08: Inservice Inspection

Procedures

ES1807.002 R6 Liquid Penetrant Examination - Solvent Removable ES1807.003 R6 Magnetic Particle Examination

MA 10.1

R2 Station Leakage Programs

MA 10.3

R2 Boric Acid Corrosion Control Program

EX1801.002 R9 Leakage Reduction Program Surveillance EX1801.006 R7 Containment Leakage Reduction Program Surveillance 54-ISI-864-01 R1 Manual phased array ultrasonic examination of weld overlaid similar and dissimilar metal welds Work Orders

WO 0700249

Repair boric acid packing leak at flow control valve 1-CS-FCV-121

WO 0705227

Repair leaking body to bonnet joint, clean and check torque

WO 0538725

Replace packing of reactor coolant seal injection valve 1-CS-V-158

WO 0703142

Replace packing in valve 1-CS-V-625

WO 0615086

Installation of vent with isolation valve on containment building spray

WO 0643870

Installation of suction and overflow piping from

TK-RWST-8

Condition Reports

CR 08-00846

Boric acid leak at packing is moist - 1-CS-V-158

CR 07-08771

Body to bonnet leak at

CS-V-150

CR 07-13845

CS-V-625 Boric acid leak reclassified to active from large dry deposit

CR 07-10652

Packing has changed leakage classification to active from dry

CR 08-05318

Dry boric acid at

RC-PCV-456A PORV and

RC-V-122

CR 08-04497

CR initiated to document results of walkdown in Mode 3

CR 08-05322

Excessive dry boric acid at

SI-V-301

CR 08-05320

Excessive dry boric acid at

RC-V-21 packing

CR 08-05318

Excessive dry boric acid at

RC-PCV-456A PORV,

RC-V-122 PORV

Attachment

CR 08-04345

Excessive dry boric acid at

CBS-V17 gland leak-off plug

CR 07-02239

Formal documentation of

MRP-139 evaluations needs to be done

Miscellaneous

Program Health Report Boric Acid Corrosion Control Program - 2008 1

st Quarter, 2007 4

th

Quarter, 2007 2

nd Quarter Examiner Qualifications Qualifications for examiners 2239, 7299, 1279 for PT- MT- UT- VT

Section 1R11: Licensed Operator Requalification

Simulator Demonstration Exam on 5/22/08

Simulator Demonstration Scenario on 5/19/08

ER2.0B, Seabrook Station State Notification Fact Sheet, Revision 30

NT-5701-3, Crew Simulator Evaluation, 5/22/08 OS1000.10, Operations at Power, Revision 5 OS1231.04, Rapid Down Power, Revision 1 OS1227.02, Steam Generator Tube Leak, Revision 12

OS1290.03, Response to a Security Event, Revision 4 OS1290.04, Response to an Airborne Security Event, Revision 0 E-0, Reactor Trip and Safety Injection, Revision 45 E-3, Steam Generator Tube Rupture, Revision 38

Section 1R12: Maintenance Rule Implementation

System Health Reports - Heater Drain System System Health Report - Solid State Protection System Seabrook System and Performance Reports Plant Engineering Guidelines, Maintenance Rule Program Monitoring Activities Plant Engineering Action Plan Register Maintenance Rule Failures Evaluated in the Condition Report System SM 7.10, Maintenance Rule Program, Revision 1 Work Orders for 2007-2008 Condition Reports for 2007-2008

Engineering Evaluation for Condition Report

200808572

Condition Reports

200808572

Section 1R13: Maintenance Risk and Emergent Work

Work Orders

0817619,

0817973,

0818508,

0816135,

0815298,

0815214,

0815228,

0815229,

0815257 Temporary Modification 08TMOD007 10

CFR 50.59 Screen Leak Repair for Valve 1-MS-V-298, 5/6/08 Engineering Evaluation:

CO-E-27-C Conderser Tube Leak, 6/6/08 Integrated Technologies Preliminary Report 13-98, 6/6/08 MA 4.14A Troubleshooting Control Form

CO-E-27C ON1040.04, Operation of the Heater Drain Pumps, Rev. 04, Chg 12

PID No. 1-HDB-20338 Adverse Condition Monitoring and Contingency Plan, 05/14/08

CR 200808615, during disassembly of the mechanical seal on 1-HD-P-31-A the carbon bushing in the stuffing box was found hard up against the rotary holder of the seal Attachment

CR 200808620, ON1040.04 has conflicting information for what seal water flow should be prior to pump start.

CR 200808532, develop an MSE to allow flexible couplings to be installed on the heater drain pump mechanical seal supply tubing.

CR 200808562, new mechanical seal stuffing box flange gasket failed while returning 1-HD-P-31-A to service post maintenance

CR 200808572, Failure Investigation Process (PI-AA-100-1002) has been initiated to investigate the cause of repeated failures of the mechanical seal on 1-HD-P-31-A. OS1234.02, "Condenser Tube or Tube Sheet Leak"

WO 0818508 SM 7.21, Revision 00, Change 01, "Condenser Waterbox Leak Response"

PI-AA-100-1002-F03, Failure Investigation Process Action Item List CD0905.07, "Seawater In-leakage" CP 3.2, Secondary Chemistry Control Program OS1000.06, Revision 6, "Power Decrease" OS1000.05, Revision 6, Change 18, "Power Increase"

CR 200808642 MPCS Alarm D4011 based on Low Lube Water Flow to

CW-P-39C

CR 200808647 Entry into OS1234.02, Condenser Tube or Tube Sheet Leak due to increased Sodium concentrations in the "C" condenser waterbox

CR 200808665 On reaching 30% power, the Power Range Nuclear Instruments gains required course adjustment.

CR 200808678 re-declared to new power level of 48% at 3 % per hour.

Section 1R15: Operability Evaluations

Condition Reports

08-04857, 08-04898, 08-05414 Technical Specifications 3.3.2, 3.4.9.3, 3.5.3.1, 3.6.2.1

OS1001.02, Draining the Reactor Coolant system for Vessel head Removal, Revision 9 OS1013.04, Residual Heat Removal System Train B Startup and Operation, Revision 12 OS1013.03, Residual Heat Removal System Train A Startup and Operation, Revision 12 OX1426.22, "Emergency Diesel Generator 1A 24 Hour Load Test and Hot Restart Surveillance." Calculation C-S-1-E-0161-establishes maximum allowable diesel fuel oil consumption rate.

CR 200808673 Prompt Operability Determination, 06/06/08 Technical Specification Bases Change 08-05, 4/28/08 Work Orders

0814321,

1641218,

0633408,

0815528, 0816375

Section 1R18: Plant Modifications

Design Change 06DCR008, Containment Sump Screens for Generic Letter 2004-02

Generic Letter 2004-02, Potential impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors FPLE Presentation on Containment Sump Strainer Modification FPLE Letter to NRC, Potential Impact of Debris blockage on Emergency Recirculation During Design Basis Accidents at Pressurized Water Reactors, 2/28/08 OX1401.02, RCS Steady State Leak Rate Calculation, Revision 4

OX1406.12, 18 Month Containment and Containment Spray Recirculation Sump Surveillance, Revision 5 Engineering Evaluation 08-003, NUKON Insulation Assessment NRC Safety Evaluation Related to NRC Generic Letter 2004-02, NEI Guidance Report WCAP 16710, Jet Impingement Testing to Determine ZOI of Nukon for Wolf Creek and Attachment Callaway Nuclear Operating Plants Foreign Print 25916, GE Hatachi 0000-0049-8050-R2, Containment Recirculation Sump Strainer System - S0100 Hydraulic Sizing Report Areva Calculation

SBC-1024, Integrated ECCS & CBS Recirculation Model Using Relap5YA

Foreign Print 59993, Sargent & Lundy Calculation 2006-00980, Debris Generation Due to LOCA within Containment for Resolution of

GL 2004-02 Foreign Print

500044, Areva Document 51-9011247-01, FPL Fleet ZOI Coatings Reduction Test Report

Section 1R19: Post Maintenance Testing

Work Orders

0640186,

0803981,

0713920,

0640851,

0640846,

0636928 WOs for B EDG Overhaul -

082117,

0821176,

0736770,

0727037,

0715982,

0724927,

08035760,

0717700,

0717730,

0715978,

0727043,

0717731,

0727067,

0727063,

0732910,

0732915,

0726996,

0724964,

0730993,

0603690,

0603686,

0736771,

0819292,

0727045,

0803485,

0729096,

0726830,

0717693 OX1456.92, Centrifugal Charging Comprehensive Pump Test

LSO569.23, MOV Spring Pack Maintenance and Testing, Revision 1 OX1405.10, Safety Injection System Cold Shutdown Valve Test, Revision 5 LN0561.03, Reserve Auxiliary Transformer Preventive Maintenance, Rev. 00, Chg. 10 OS1048.13. Vital Buss 11A Operation, Rev. 00, Chg. 03

Section 1R20: Refueling and Outage Activities

Control Room Narrative Logs Main Control board and MPCS Plant Parameter Displays and Trends Engineering Evaluation EE08-004, OR12 Shutdown Safety Evaluation OS1000.01, Heatup from Cold Shutdown to Hot Standby, Revision 13

OS1000.02, Plant Startup from Hot Standby to Minimum Load, Revision 8 OS1000.03, Plant Shutdown from Minimum Load to Hot Standby, Revision 5 OS1000.04, Plant Cooldown from Hot Standby to Cold Shutdown, Revision 8 OS1000.05, Power Increase, Revision 6 OS1000.06, Power Decrease, Revision 6

OS1000.07, Approach to Critical, Revision 6 OS1000.09, Refueling Operation, Revision 7 OS1000.15, Refueling Outage Cooldown, Revision 1 OS1001.01, Heatup from Cold Shutdown to Hot Standby, Revision 13 OS1001.11, Reactor Coolant System Shutdown Level, Revision 1

OS1007.01, Automatic and Manual Rod Control, Revision 7 OS1012.03, Primary Component Cooling Water Loop A Operation, Revision 13 OS1012.04, Primary Component Cooling Water Loop B Operation, Revision 11 OS1013.03, Residual Heat Removal System Train A Startup and Operation, Revision 12 OS1013.04, Residual Heat Removal System Train B Startup and Operation, Revision 12 OS1014.02, Operation of Spent Fuel Cooling and Purification System, Revision 8 OS1015.05, Fuel Transfer System and Upender Operation, Revision 7

OS1015.07, Spent Fuel Bridge Assembly Operation, Revision 7 OS1015.18, Setting Containment Integrity for Mode IV Entry, Revision 5, Change 29 OS1036.01, Aligning the Emergency Feedwater System For Automatic Initiation, Revision 8 ON1031.02, Starting and Phasing the Turbine Generator, Revision 6 OX1401.02, RCS Steady State Leak Rate Calculation, Form B, Revision 4

Attachment OX1406.12, 18 Month Containment and Containment Spray Recirculation Sump Surveillance, Revision 5, Change 12 RS0721, Refueling Administrative Control, Revision 4 RS1735 Form A, Estimated Critical Position Data and Analysis, Revision 4

Reactor Engineering Operating Recommendation 08-REOR-008, Guidance for Beginning of Cycle 13 Power Ascension Open Condition Reports and Actions with Mode Restrictions Mode Change Report Mode 6 to Mode 5 Mode change Report Mode 5 to Mode 4

Mode Change Report for Modes 3, 2, 1 License Amendment No. 117, Setpoint Change for Reactor Trip system Interlock P-9, 3/27/08 License Amendment No. 118, Nuclear Instrumentation Surveillance Requirements, 4/29/08 Condition Report

200807924,

200805414,

200805758,

200805751 Clearances MT018, MT003, MT700-16A, MT004, MT006-05, MT005-5A,5B,5C FPLE Letter

SBK-L-08074, Preliminary OR12 Structural Weld Overlay Examination Results, 5/12/08 FPLE Letter

SBK-L-08103, Notification of Completed Actions and Commitments Addressed in Confirmatory Action Letter In Regard to Alloy 82/182 Butt Welds in the Pressurizer at Seabrook Unit No. 1 SBK

QR 08-0027, Assessment of

NP-910, Readiness for Plant Operations for

OR-12

NP-910, Plant Readiness for Operations, 5/3/08

Work Orders

0640089,

0817548 Refuel 12 Clearance Tag List for MT019-01 Maintenance Manual, Section MA 4.2, Equipment Tagging and Isolation, Rev 20 Chg 1 Maintenance Manual, Section MA 3.4, Foreign Material Exclusion, Rev 11 Chg 13 Seabrook Station Administrative Procedure, Condition Reports, OE 3.6, Rev 13

IS1680.909, Train A SSPS Power Supply Replacement (Logic Bay), Rev 0 Chg 1 IS1680.914, Solid State Protection System Train A Power Supply Test, Rev 2 Chg 2 IS1680.915, Solid State Protection System Train B Power Supply Test, Rev 2 Chg 2 Seabrook Updated FSAR, Section 7.2 Reactor Trip Protection Seabrook Updated FSAR, Section 9.1.3, Spent Fuel Pool Cooling and Cleanup System Seabrook Updated FSAR, Section 9.2 Component Cooling Water Work Order

0638070, SG Secondary Side Handhole cover Removal and Installation Work Order

0639798, SSPS Train A Power Supply Test Work Order

0639800, SSPS Train B Power Supply Test Work Order

0711876, Install/Test New Power Supplies in SSPS Train A

CP-12

Work Order

0711877, Install/Test New Power Supplies in SSPS Train B

CP-13 Work Order

0713921, Limitorque MOV Inspection Work Order

0814297, Operability Testing of IST Valves Work Order

0815238, Analyze Spare New SSPS Power Supply to Determine Cause of

Failure of OVPD OX1456.81, Operability Testing of IST Valves, Rev 6 Chg 2 Technical Specification 3/4.9.4 Containment Penetrations (Refueling Operations)

Control of Heavy Loads

UFSAR 9.1.5, Overhead Heavy Load Handling System, Revision 12

NRC

EGM 07-006, Enforcement Discretion for Heavy Load Handling Activities, 9/28/07 OpESS FY2007-03, Crane and Heavy Lift Inspection, Supplemental Guidance for IP71111.20

Attachment AES Calculations

SBK-12905-M01 through M04, Evaluation of a Postulated Reactor Head Drop Accident, 3/24/08 Crane Qualification Training Records NEI Letter, Guidelines for Reactor Vessel Head Drop Analyses, 1/15/08

AES Report Seabrook Reactor Vessel Head Drop Analysis Methodology and Acceptance Criteria, February 29, 2008 10CFR50.59 Screen 08-162 for 08MSE057 Maintenance Support Evaluation 08MSE057,

NUREG-0612 Response - Reactor Vessel Head Drop Analysis

MS0540.10, Reactor Vessel Head Removal and Storage, Revision 6, Change 11 Arrangement Drawings 1-NHY-805056, 1465E89, 1-NHY-203000 Drawing 1-NHY-805272, Safe Load Paths Work Oder

0638067, Polar Gantry Crane Refueling Outage Inspection MS0504.16, Reactor Vessel Head Installation, Revision 6 MN0534.01, Polar Gantry Crane Inspection Work Orders

0735647,

0639522,

0611152,

0632454,

0638068,

0638352, 0638380

Section 1R22: Surveillance Testing

Work Orders

0640186,

0720334,

0640252

OX1445.92, Centrifugal Charging Pump Comprehensive Pump Test, Revision 0 OX1426.19/20, Diesel Generator 1A/1B 18 Month Operability and Engineered Safeguards Pump and Valve Response Time Testing Surveillance, Revision 4/3 EX1803.003, Reactor Containment Type B and C Leakage Rate Tests, Revision 6 EX1803.001, Reactor Containment Integrated Leakage Rate Test - Type A, Revision 4 Technical Specification 6.15, Containment Leakage Rate Test Program, Amendment 108 Reactor Containment Building Integrated Leakage Rate Test Report, 1/19/93

Control Room Narrative Logs Main Control board and MPCS Plant Parameter Displays and Trends ANSI/ANS-56.8-1994, Containment System leakage Testing Requirements

BN-TOP-1, Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants,

Revision 1, 11/1/72 ILRT Letter

GCVWL-080426A, ILRT Data Summary, 4/26/08 OX1426.22, Revision 01, Change 06, "Emergency Diesel Generator 1A 24 Hour Load Test and Hot Restart Surveillance." OX1426.16, DG 1A Tech Spec Action Statement Surveillance, Rev. 04, Chg. 17 DG 1A operations logs (9 pages)

DG 1A Log Sheet (17 pages) DG 1A Log Sheet (39 pages) CX0901.22, Revision 14, "Diesel Generator fuel Oil Tank Surveillance"

CR 08-08485 Diesel fuel intended for the emergency diesel generator was not accepted base on inconsistent Bill of Lading and product code information.

CR 08-07465 for DG 1A jacket water keep-warm heater control circuit failure.

WO 0817745 for DG 1A jacket water keep-warm heater control circuit failure Emergency Diesel Fuel Oil Log,

CHL-039 for

WO 0818186 and 0818187

Section 2OS1: Access Control to Radiologically Significant Areas

Procedures

HD0958.03, Rev 23 Personnel Survey and Decontamination Techniques Attachment HN0958.13, Rev 26 Generation and Control of Radiation Work Permits HD0958.17, Rev 12 Performance of Routine Radiological Surveys HD0958.19, Rev 27 Evaluation of Dosimetry Abnormalities HN0958.25, Rev 26 High Radiation Area Controls

HD0958.30, Rev 23 Inventory and Control of Locked or Very High Radiation Area Keys and Locksets HD0958.48, Rev 02 Health Physics Job Coverage Using Remote Monitoring HD0958.51, Rev 00 Health Physics Issuance of Stop Work Orders HD0992.02, Rev 28 Issuance and Control of Personnel Monitoring Devices

HN0958.30, Rev 23 Inventory and Control of Locked or Very High Radiation Area Keys and Locksets HN0958.39, Rev 04 Multi-Badge Control & Exposure Tracking RP 2.1, Rev 18 General Radiation Worker Instruction and Responsibilities RP 3.1, Rev 18 Radiological Qualification Requirements RP 4.1, Rev 19 Requirements for Issuing Personnel Dosimetry RP 5.1, Rev 16 Annual Occupational Exposure Control and Increased Radiation Exposure Approval RP 9.1, Rev 21 RCA Access/Egress Requirements

RP 13.1, Rev 20 Radiological Controls for Materials

RP 13.2, Rev 5 Storage of Highly Radioactive Material in the Reactor Cavity or Spent Fuel Pool

RP 15.1, Rev 18 Job Pre-Planning and Review for Radiation Exposure Control

RP 15.2, Rev 09 ALARA Recommendations

RP 15.4, Rev 11 Use and Control of Temporary Shielding Quality Assurance Reports

Daily Quality Summary Reports for the period January 21, 2008 through April14, 2008

Radiation Protection/ Process Control,/RadWaste Programs Audit (SBK-08-01)

Condition Reports

08-01812, 08-02733, 08-04975, 08-05505, 08-04951, 08-04562, 08-02299, 08-01672, 08-

03476, 08-04969, 08-04657, 08-04753, 08-01577, 08-04881,

Miscellaneous

Health Physics Study/Technical Information Document (HPSTID 08-0036), Evaluation of Effective Dose Equivalent for Pressurizer Weld Overlay Job Selected Temporary Shielding Requests

Reactor Coolant Chemistry Post Shutdown Data Reactor Coolant System Piping Dose Rates Post Shutdown

ALARA Reviews/Radiation Work Permits

AR 08-01, OR12 RV Disassembly & Reassembly/

RWP 08-025

AR 08-02, OR12 Steam Generator Secondary Side Maintenance & Inspections/RWP 08-033

AR 08-03, OR12 Inservice Inspection/

RWP 08-035

AR 08-04, OR12 Cavity Decon/RWP 08-08-028

AR 08-05, OR12 MOV Testing/Preventative Maintenance & Repair/

RWP 08-08-037

AR 08-06, OR12 Valve Maintenance/

RWP 08-08-036

AR 08-07, Fuel Handling/

RWP 08-08-026

Attachment

AR 08-09, OR12 RCP Seals and Motor Maintenance/

RWP 08-08-034

AR 08-10, OR12 Scaffold/

RWP 08-08-038

AR 08-11, OR12 Insulation/

RWP 08-08-045 AR08-12,

Pressurizer Nozzle Weld Overlay and NDE Project/

RWP 08-048 & 08-056

AR 08-13, ECCS Sump Modification/ RWP 08-047

Section 4OA2: Identification and Resolution of Problems

CAP Quarterly Trend Reports for Second, Third and Fourth Quarter 2007 CAP Quarterly Trend Report for first Quarter 2008

Condition Reports

for 2007-2008

Section 4OA3: Event Follow-up

Documents

Reviewed:

MA 4.2, Equipment Tagging and Isolation, Revision 20 MA 4.5, Configuration Control, Revision 13 OS1090.05, Configuration Control, Revision 5

Condition Reports

200806229,

200806265,

200806278,

200806270,

200806260,

200806265,

200713999,

200714273,

200714536,

200806409 Technical Specification

6.7.1.a Regulatory Guide 1.33, Appendix A Control Room Logs Mid Shift and Day Shift 04/20/08

Clearance MT005-05A, 05B and 05C Work Order

0518410,

0814050 ER 1.1, Classification of Emergencies, Revision 45 OS1002.01, filling and Venting the Letdown, Charging and Seal Injection and Excess Letdown Portions of the CVCS, Revision 7

OS1002.02, Operation of Letdown, Charging and Seal Injection, Revision 14 MS01519, Grinnell Diaphragm Valve Maintenance, Revision 5 OS1015.10, Refueling Canal and Cavity Drain, Revision 6 OS1200.00, Response to Fire or Fire Alarm Actuation, Revision 14 OS1231.04, Rapid Downpower, Revision 1

ON1038.08, Draining and Filling condenser Waterboxes, Revision 5 OPMM, Rev 84, Section 1.19 Standing Order 08-009, Dated 04/21/08 Alarm Response Procedure

UA-8334 A-5, PAB Sump A Level Hi PID 1-CS-B20723, Chemical & Volume Control System, Purification Rev 19

PID 1-CS-B20727, Chemical & Volume Control System, Thermal Regeneration Rev 8 Licensed Operator Continuing Training Package L5076C0810 (for

CR 07-14273, Action 01)

Attachment

Section 4OA5: Other Activities

Inspection Results for

TI 2515/172, RCS Dissimilar Metal Butt Welds

Condition Reports

CR 08-05680

Tracking of resolution of flaws found on "D" safety nozzle overlay

CR 08-05949

Pressurizer surge line safe end indication

CR 08-05958

UT indication 1/2" from toe on safe end side of pressurizer surge nozzle

CR 08-06088

Evaluation on design qualification of relief "B" nozzle UT indication

CR 08-05180

Resolution of overlay contour and as-design for safety "D" nozzle

CR 08-05775

Evaluation of incorrect calibration standard for UT of spray nozzle overlay

CR 08-05793

Evaluation of PT indications in existing dissimilar metal weld of "A" nozzle

CR 08-05795

Two areas of indications found in nozzle "D" by UT

CR 08-03183

Determine future pressurizer weld overlay examinations and frequency

CR 08-03139

Revise reactor coolant system (RCS) materials degradation management reference, Alloy 600 program Examination Reports

A-SWOL-DS01 Phased array indication data sheet, weld

RC-E-10-A-SWOL, "A" nozzle A-SWOL-PS01 Weld overlay indication plot sheet, "A" nozzle, safety relief

B-SWOL-DS01 Phased array indication data sheet, weld

RC-E-10-B-SWOL, "B" nozzle B-SWOL-PS01 Weld overlay indication plot sheet, "B" nozzle, power operated relief C-SWOL-DS01 Phased array indication data sheet, weld

RC-E-10-C-SWOL, "C" nozzle C-SWOL-PS01 Weld overlay indication plot sheet, "C" nozzle, safety relief D-SWOL-DS02 Phased array indication data sheet, weld

RC-E-10-D-SWOL, "D" nozzle

D-SWOL-PS02 Weld overlay indication plot sheet, "D" nozzle, safety relief

SP-SWOL-DS01 Phased array indication data sheet, weld

RC-E-10-SP-SWOL spray

SP-SWOL-PS01 Weld overlay indication plot sheet, spray nozzle S-SWOL-DS01 Phased array indication date sheet, weld

RC-E-10-S-SWOL surge

S-SWOL-PS01 Weld overlay indication plot sheet, surge nozzle

08-02-028

Penetrant test of tee to pipe, weld

RH 0179-01-01 08-02-002

Penetrant test of elbow to pipe, weld

CBS 1213-02-03 08-03-020

Magnetic particle test of elbow to pipe, weld

MS 4002-02-09 08-03-002

Magnetic Particle test of valve to pipe, weld

FW 4606-03-06 08-03-022

Magnetic Particle test of pipe to elbow, weld

MS 4002-36-06

Work Orders

WO 0700249

Repair boric acid leak at packing, flow control valve 1-CS-FCV-121

WO 0705227

Clean boric acid crystals and evaluate leak source, valve 1-CS-V-150

WO 0538725

Boric acid crystals at packing, evaluate and repair, valve 1-CS-V-158

WO 0703142

Repair body to bonnet and packing leakage, valve 1-CS-V-625

Welding Procedures (WP) and Procedure Qualification Records (PQ)

WP3/8/F43OLTBSCa3-003 Machine Temper Bead Overlay, Gas Tungsten Arc Welding (GTAW) of P3 to P8 using F43 filler metal and PQ7164, 7213, 7280 and 7281 WP3/8/F430LTBSCa3-002

Machine Temper Bead Overlay, GTAW, P3 to P8 using F43 filler metal, provides for orbital weld progression Attachment WP8/8F6AW3-07 Machine GTAW of P8 to P8 using F6 filler metal and PQ7062 WP8/8F6AW1

Manual GTAW of P8 to P8 using F6 filler metal and PQ7037 and 7038

Drawings

1-NHY-801213 ISI Containment Spray System - Line No. 1213

1-NHY-800179 ISI RHR System - Line No. 179 1-NHY-202303 ISI Main Steam System - Line No. 4002

1-NHY-202396 ISI Feedwater System - Line No. 4606

1-NHY-804002 ISI Main Steam System Loop 3 - Line No. 4002 8020573D R001 Pressurizer Safety & Relief Nozzle Weld Overlay Design Input

Miscellaneous

Performance Demonstration Initiative Procedure/Personnel Performance Qualification Records

DCN 06DCR012-01 Design change notice, pressurizer nozzle weld overlay installation

06DCR012

Applicability determination and 50.59 screen for overlay Installation Temporary Instruction 2515/166 - Pressurized Water Reactor Containment Sump Blockage (NRC Generic Letter 2004-02)

CFR 50.59 Screens

07-200, Containment Sump Screens for

GL 2004-02, Rev. 0

Calculations

2006-09080, Transmittal of Final Seabrook Debris Transport Calculation 2006-09080, Rev. 3 4.3.05.10F, CBS Hydraulic Analysis, Rev. 9 32-5050140-00, Seabrook Station integrated ECCS & CBS Recirculation Model Using RELAP, dated 12/09/2004 C-S-1-83814, Seabrook Post Accident Chemical Product Formation, Rev. 0

GE-0000-0080-4405, Seabrook Filter Bypass Analysis, Rev. 0

GE-NE-0000-0049-8050, Containment Recirculation Sump Passive Strainer System - S0100

Hydraulic Sizing Report Seabrook Nuclear Power Plant, Rev. 2

GE-NE-0000-0067-0140, CBS Containment Debris Interceptors Transport Analysis, Rev. 1

Condition Reports

(* denotes NRC identified during this inspection) 08-06551*

Drawings

9763-F-805051, Containment Structure Elevation Plan General Arrangement 9763-F-805055, Containment Structure Elevation Plan Elevation "A-A," "B-B," and "C-C" General Arrangement 9763-F-805056, Containment Structure Elevation "D-D," "E-E," and "F-F" Attachment Modifications

06DCR008, Containment Sump Screens for

GL 2004-02, dated Rev. 0

Procedures

OX1406.12, 18 Month Containment and Containment Spray Recirculation Sump Surveillance,

Rev. 5, Chg. 10 OS1015.18, Setting Containment Integrity for Mode IV Entry, Rev. 5, Chg. 28

Miscellaneous

Letter from FPL to U.S. NRC: Supplemental Response to NRC Generic Letter 2004-002, dated 02/28/2008 Letter from FPL to U.S. NRC: Supplement to Response to NRC Generic Letter 2004-002, dated 01/27/2006 Letter from FPL to U.S. NRC: Response to NRC Generic Letter 2004-002, dated 09/01/2005 Specification S-S-1-E-1047, Protective Coatings for Service Level I Applications Inside Containment Building, Rev 0 UFSAR Section 5.4, Component and Subsystem Design UFSAR Section 6.3, Emergency Core Cooling Systems UFSAR Section 9.2, Water Systems UFSAR Section 15.6, Decrease in Reactor Coolant Inventory U.S. NRC Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors

Section 4OA7: Licensee-Identified Violations

Condition Report

200801388 UFSAR Section 1.8, Conformance to Regulatory Guides, Revision 12

FPLE Letter to NRC dated February 22, 2008, Experience Assessment Root Cause Analysis for

CR 08-01388, LOIT Candidate Screening Regulatory Guide 1.8, Qualification and Training of Personnel for Nuclear Power Plants, Revision 3

LIST OF ACRONYMS

ADAMS Agency-wide Documents Access and Management System

ALARA As Low As Is Reasonable Achievable

AR [[]]

ALARA Reviews

ASCA Advanced Scale Conditioning Agent

ASME American Society of Mechanical Engineers

CL Cold Leg

DCR Design Change Request

ED Electronic Dosimeter

EDG Emergency Diesel Generator

FME Foreign Material Exclusion

FPLE Florida Power & Light Energy

FW Feedwater

GL Generic Letter

Attachment

GSI Generic Safety Issue

GTAW Gas Tungsten Arc Welding

HL Hot Leg

HRA High Radiation Areas

IMC Inspection Manual Chapter

ISI In-service Inspection

LER Licensee Event Reports

LHRA Locked High Radiation Areas

MPCS Main Plant Computer System

MRFF Maintenance Rule Functional Failure

MRP Materials Reliability Program

MS Main Steam

MT Magnetic Particle Test

NCV Non-Cited Violation

NDE Non-Destructive Examination

NEI Nuclear Energy Institute

NRC U.S. Nuclear Regulatory Commission

NRR Nuclear Reactor Regulation

PAB Primary Auxiliary Building

PARS Publicly Available Records

PDI Performance Demonstration Initiative

PMT Post-Maintenance Testing

PQR Procedure Qualification Record

PT Penetrant Test

PWR Pressurized Water Reactor

RCA Radiological Controlled Area

RCS Reactor Coolant System

RMW Reactor makeup water

RV Reactor Vessel

RWP Radiation Work Permit

SDP Significance Determination Process

SFP Spent Fuel Pool

SG Steam Generator

SRA Senior Reactor Analyst

TI Temporary Instruction

TS Technical Specifications

UFSAR Updated Final Safety Analysis Report

UT Ultrasonic Testing

VHRA Very High Radiation Areas

VT Visual Test

WO Work Order

WPS Weld Procedure Specification

B-1 Attachment

ATTACHMENT B

TI 172 Documentation Questions for Seabrook Station

Introduction

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 (RCS) containing Alloy 600/82/182. The TI requires documentation of specific questions in an inspection report. The questions and responses are included in this Attachment B.

In summary Seabrook Station has four (4) six (6) inch pressurizer safety relief nozzles, one (1)

fourteen (14) inch surge line nozzle, one (1) four (4) inch spray nozzle, (4) twenty nine (29) inch reactor vessel hot leg (HL) outlet nozzles and (4) twenty seven and one-half (27.5) inch reactor vessel cold leg (CL) inlet nozzles which are MRP-139 applicable Alloy 600/82/182 dissimilar metal welds. Seabrook Station has submitted a proposed alternative to the ASME Code 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, 07/03/2007) and supplement (SBK-L-08022, 02/13/2008) were approved on April 1, 2008, by

NRC Staff. a. For

MRP-139 baseline inspections:

Qa1. Have the baseline inspections been performed or are they scheduled to be performed in accordance with

MRP -139 guidance?

A. No, the baseline inspections required by

MRP -139 were not completed within the prescribed time frame. The baseline

UT inspection of six (6) dissimilar metal butt welds (

DMBW) within the scope of MRP-139 were deferred from the required implementation

date of December 31, 2007 to allow for the application of a full structural weld overlay during the Spring 2008 outage (OR12). The full structural weld overlay was applied to the pressurizer surge, spray and four (4) safety nozzles during OR12. A baseline inspection of the six weld overlays was completed during this outage. This baseline inspection consisted of a liquid penetrant surface examination and a volumetric manual

Performance Demonstration Initiative (PDI) Qualified phased array ultrasonic examination. The eight (8) reactor vessel

RCS welds (4 hot leg and 4 cold leg) are scheduled to be

PDI qualified

UT examined during the Fall 2009 (

OR13) outage.

B-2 Attachment

Qa2. Is the licensee planning to take any deviations from the MRP-139 baseline inspection requirements of MRP-139? If so, what deviations are planned and what is the general basis for the deviation? If inspectors determine that a licensee is planning to deviate from any MRP-139

baseline inspection requirements,

NRR [[should be informed by email as soon as possible. A. Yes. A deviation was submitted to extend the implementation of the required baseline inspections from the date of December 31, 2007 []]

MRP 139, 1.2(2)] for approximately three (3) months to facilitate the application of a full structural weld overlay of six pressurizer dissimilar welds. In lieu of baseline UT inspections, full

structural weld overlays were applied to dissimilar welds in the surge and spray lines and the four safety/relief nozzles. The full structural weld overlays were applied in accordance with Alternative Repair Technique request submitted on July 3, 2007 and approved by

NRC on April 1, 2008. The four reactor vessel

RCS hot leg (29") and four cold leg (27.5") welds will be

UT examined using

PDI qualified procedures and personnel as required by MRP-139, during the Fall 2009 outage. The basis for the deviation was the determination that the six welds were not inspectable to achieve 90%

(or greater) examination volume prescribed by

MRP -139 because of obstructions and the current weld contour.

NRC was notified of this deviation on March 6, 2006 by Seabrook letter

SBK-L-06044. The Nuclear Energy Institute (NEI) 03-08 process was followed for this deviation.

b. For each examination inspected, was the activity:

Qb1. Performed in accordance with the examination guidelines in

MRP -139 Section 5.1 for unmitigated welds or mechanical stress improved welds and consistent with

NRC staff relief request authorization for weld overlaid welds? A. Yes. The overlay activity of the six previously identified welds were applied and examined in accordance with the examination guidelines in

MRP-139 and the relief request authorization. The relief request authorization permitted the application of a full structural weld overlay with subsequent liquid penetrant surface examination and volumetric PDI qualified phased array ultrasonic examination of the weld overlay.

Mechanical stress improvement was not used on any dissimilar weld.

Qb2. Performed by qualified personnel? (Briefly describe the personnel training/qualification process used by the licensee for this activity.)

A. Yes. The examinations were performed by personnel qualified to the requirements of

ASME Section XI, Appendix

VIII. Procedures and personnel were qualified in the

PDI program for the manual phased array ultrasonic examination of weld overlays on similar and dissimilar metal welds.

Qb3. Performed such that deficiencies were identified, dispositioned, and resolved? A. Yes. Indications identified in the ultrasonic examination were evaluated for

relevance, characterized and entered into FPLE's corrective action program for disposition and resolution.

B-3 Attachment

c. For each weld overlay inspected, was the activity:

Qc1. Performed in accordance with

ASME Code welding requirements and consistent with

NRC staff relief request authorizations? Has the licensee submitted a relief request and

obtained

NRR staff authorization to install the weld overlays? A. Yes. The deposit of the weld overlays was performed in accordance with the

ASME Code requirements (Section

IX and

XI) using qualified procedures and welders. Weld overlay of the six dissimilar metal welds was authorized by NRR in their approval dated April 2, 2008, for Seabrook Station to perform a full structural weld overlay on the surge,

spray and four safety/relief welds.

Qc2. Performed by qualified personnel? (Briefly describe the personnel training/qualification process used by the licensee for this activity.)

A. Yes. The welders were qualified to

ASME Section IX and personnel inspecting the

completed weld overlays were qualified in accordance with

ASME Section

XI , Appendix

VIII and

PDI qualified for manual phased array ultrasonic examination. A representative dissimilar weld mock-up was used for training test examiners.

Qc3. Performed such that deficiencies were identified, dispositioned, and resolved?

A. Yes. Indications identified in the ultrasonic

PDI-UT examination were evaluated for relevance, characterized and entered into FPLE's corrective action program for disposition and resolution.

d. For each mechanical stress improvement used by the licensee during the outage, was the activity performed in accordance with a documented qualification report for stress improvement processes and in accordance with demonstrated procedures? Specifically:

Qd1. Are the nozzle, weld, safe end, and pipe configurations, as applicable, consistent with the configuration addressed in the SI qualification report? A. N/A, mechanical stress improvement was not used.

Qd2. Does the SI qualification report address the location radial loading is applied, the applied load, and the effect that plastic deformation of the pipe configuration may have on the ability to conduct volumetric examinations? A. N/A

Qd3. Do the licensee

=s inspection procedure records document that a volumetric examination per the

ASME Code, Section

XI , Appendix

VIII was performed prior to and after the application of the

SI? A. N/A

Qd4. Does the

SI qualification report address limiting flaw sizes that may be found during pre-

SI and post-SI inspections and that any flaws identified during the volumetric examination are to be within the limiting flaw sizes established by the SI qualification report. A. N/A

B-4 Attachment Qd5. Performed such that deficiencies were identified, dispositioned, and resolved?

A. N/A e. For the inservice inspection program

Qe1. Has the licensee prepared an MRP-139 inservice inspection (ISI) program? If not, briefly summarize the licensee

=s basis for not having a documented program and when the licensee plans to complete preparation of the program.

A. Yes.

FPLE has an MRP-139

ISI program, which is implemented through the Reactor Coolant System Materials Degradation Management Program and is separate from the

ASME Section

XI [[]]

ISI program. In the interim, the MRP-139 inservice inspection program is implemented through the existing Alloy 600 Aging Management Program that contains the strategy for all alloy 600/82/182 pressure boundary butt weld locations at Seabrook

Station. This plan includes inspections, examination schedules, mitigation and repair/replacement activities. Welds will be added to the Section

XI [[]]

ISI program when mitigation or repair/replacement activities have been completed.

Qe2. In the MRP-139 inservice inspection program, are the welds appropriately categorized in

accordance with

MRP -139? If any welds are not appropriately categorized, briefly explain the discrepancies.

A. Yes. All welds are categorized per

MRP-139 requirements as applicable.

Qe3. In the MRP-139 inservice inspection program, are there inservice inspection frequencies,

which may differ between the first and second 10-year intervals after the

MRP -139 baseline inspection, consistent with the inservice inspection frequencies called for by MRP-139?

A. All

MRP-139 applicable welds are scheduled either for mitigation and/or inspection prior to the end of the current 10-year ISI inspection interval which ends in August 2010.

Qe4. If any welds are categorized as H or I, briefly explain the licensee

=s basis for the categorization and the licensee

=s plans for addressing potential

PWSCC. A. Four welds are categorized as Category H, (

RCS HL,

RV outlet) and four welds are categorized Category I (

RCS CL,

RV [[inlet) and are 27.5" and 29.0" diameter, respectively. These eight dissimilar welds have been placed in category "H" and "I" based on their being made of non-resistant materials and outside surface conditions and access obstructions limited the exam coverage to less than 90% of the examination volume.]]

PDI qualified procedures and personnel were not available for use at the time

of the last

UT examination. No rejectable indications were identified during the previous

ASME Section

XI [[]]

UT [[exam. At this time, no plan is in place for application of any currently used mitigation methods. However, these eight welds are scheduled to be examined during the next refuel outage (OR13) that will occur in October of 2009. The welds will be examined using a]]

PDI qualified procedure implemented by

PDI qualified

personnel. At that time, it is expected that the category for the hot leg welds will change from H to D and the cold leg welds will change from I to E.

B-5 Attachment

Qe5. If the licensee is planning to take deviations from the inservice inspection requirements of

MRP -139, what are the deviations and what are the general bases for the deviations? Was the

NEI 03-08 process for filing deviations followed? A. No deviations are currently planned for any

ISI of the welds to MRP-139 at Seabrook Station.