IR 05000443/2006005

IR 05000443-06-005, 10/1/2006-12/31/2006; Seabrook Station, Unit 1; Refueling and Outage Activities

ML070180574
Person / Time
Site:	Seabrook
Issue date:	01/17/2007
From:	Paul Krohn NRC/RGN-I/DRP/PB6
To:	St.Pierre G Florida Power & Light Energy Seabrook
Krohn P, RI/DRP/PB6/610-337-5120
References
IR-06-005
Download: ML070180574 (45)

Text

ary 17, 2007

SUBJECT:

SEABROOK STATION - NRC INTEGRATED INSPECTION REPORT 05000443/2006005

Dear Mr. St. Pierre:

On December 31, 2006, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at the Seabrook Nuclear Power Station. The enclosed report documents the inspection findings which were discussed on January 9, 2007, with Mr. G. S and other members of your staff.

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

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

This report documents one NRC-identified finding and one self-revealing finding of very low safety significance (Green). These findings were determined to involve violations of NRC requirements. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs), in accordance with Section VI.A.1 of the NRC Enforcement Policy. 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, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at Seabrook.

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

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

Sincerely,

/RA/

Paul G. Krohn, Chief Projects Branch 6 Division of Reactor Projects Docket No. 50-443 License No: NPF-86 Enclosure: Inspection Report No. 05000443/2006005 w/ Attachments: Supplemental Information and Temporary Instruction 2515/150, Reactor Pressure Vessel Head and Vessel Head Penetration Nozzles cc w/encl:

J. A. Stall, FPL Senior Vice President, Nuclear & CNO M. Warner, Vice President, Nuclear Support Services R. S. Kundalkar, FPL Vice President, Nuclear Engineering M. Mashhadi, Senior Attorney, Florida Power & Light Company M. S. Ross, Managing Attorney, Florida Power & Light Company J. M. Peschel, Manager, Regulatory Programs M. Kiley, Plant General Manager, Seabrook Station J. Dent, Assistant Plant Manager K. Wright, Manager, Nuclear Training, Seabrook Station R. Poole, FEMA, Region-I Office of the Attorney General, Commonwealth of Massachusetts K. Ayotte, Attorney General, State of New Hampshire O. Fitch, Deputy Attorney General, State of New Hampshire D. Harnish, 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 New Hampshire J. Giarrusso, MEMA, Commonwealth of Massachusetts D. O'Dowd, Administrator, Radiological Health Section, DPHS, DHHS, State of New Hampshire Administrator, Bureau of Radiological Health, State of New Hampshire J. Roy, Director of Operations, Massachusetts Municipal Wholesale Electric Company T. Crimmins, Polestar Applied Technology R. Backus, Esquire, Backus, Meyer and Solomon, New Hampshire 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

Mr. Gene S

SUMMARY OF FINDINGS

IR 05000443/2006005; 10/1/2006-12/31/2006; Seabrook Station, Unit 1; Refueling and Outage

Activities.

The report covered a 13-week period of inspection by resident inspectors, regional inspectors providing resident inspection support, an announced inspection by a regional health physics inspector, and an announced inspection by a reactor inspector. Two Green non-cited violations (NCVs) were identified. The significance of most findings is indicated by their color (Green,

White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

Cornerstone: Mitigating Systems

C

Green.

A self-revealing non-cited violation of Technical Specification 6.7.1.a,

Procedures and Programs was identified by the inspectors. On October 25, 2006, during a cold shutdown, operators inadvertently performed a 500-gallon dilution instead of a planned blended makeup to the reactor coolant system.

Seabrook determined the root cause of this event was a loss of configuration control of the boric acid storage system due to a lack of procedure use in accordance with established standards. This resulted in isolating the normal flow path from the boric acid storage system for boric acid additions and blended makeups. Seabrook developed numerous corrective actions to address the root and contributing causes with the focus on improving the station adherence to standards.

The finding is more than minor because if left uncorrected it would become a more significant safety concern. Specifically, if the dilution occurred while the plan was on-line, this would have resulted in a more significant reactivity change and potential overpower condition. The finding was determined to be of very low safety significance (Green) since the reactivity change did not result in exceeding the Technical Specification shutdown margin requirements. This finding has a cross-cutting aspect in the area of human performance (sub-category work practices) because personnel did not follow procedural compliance standards.

(Section 1R20.3)

Cornerstone: Barrier Integrity

Green.

The inspectors identified a non-cited violation of Technical Specification 6.7.1.a, Procedures and Programs. On October 11, 2006, Seabrook failed to adequately establish and implement procedural controls for a heavy load lift, which resulted in a reactor coolant floor plug passing over an open, partially-fueled reactor vessel. A stand-down of the affected rigging personnel iii immediately occurred and subsequent moves were conducted within the safe load path.

The finding was more than minor because it could be reasonably viewed as a precursor to a significant event because a portion of the heavy load traveled over the reactor vessel that contained irradiated fuel and the reactor vessel head was removed for refueling activities. This finding was not suitable for a significance determination process evaluation, but has been reviewed by NRC management and was determined to be a finding of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance (sub-category resources) because Seabrook did not have complete and accurate procedural controls to assure safety of the heavy load lift. (Section 1R20.2)

Licensee-Identified Violations

None.

iv

REPORT DETAILS

Summary of Plant Status

The plant began the inspection period in its eleventh refueling outage (OR11). On November 3, 2006, the unit entered Mode 2, achieved criticality, and completed low power physics testing.

The unit was shutdown while turbine generator re-assembly continued. On November 6, 2006, the unit re-entered Mode 2 and achieved criticality. On November 9, 2006, the main generator breaker was closed briefly for static exciter testing and turbine overspeed trip tests. On November 10, 2006, the main generator breaker was closed completing OR11. On November 13, 2006, the unit reached full power operation. The unit operated at or near full power for the remainder of the period, except for a brief four percent power reduction on November 27, 2006, which was based on a request from the Regional Independent System Operator.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01 - Two Samples)

a. Inspection Scope

The inspectors reviewed Seabrooks preparation for adverse weather relative to the freeze protection of safety-related structures, systems, and components from cold weather and winter storms. The inspectors sampled adverse weather preparations on various portions of the emergency feedwater and main steam systems. The systems were reviewed to determine the adequacy of Seabrooks preparation for and ability to respond to cold weather and winter storms. The inspectors also conducted site walkdowns to verify site readiness for adverse weather including the impact of snow removal activities.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

.1 Full System Walkdown - Residual Heat Removal System (71111.04 - One Sample)

a. Inspection Scope

The inspectors conducted a detailed review of the alignment and conditions of the residual heat removal system. The inspectors performed a walkdown to verify the system alignment was maintained in accordance with system drawings and procedures.

Control room indications were verified to be appropriate and consistent with Technical Specification requirements and the Update Final Safety Analysis Report (UFSAR). The inspectors reviewed and evaluated the potential impact on system operation from open work orders, condition reports, tagged equipment, and reduced inventory operations.

System health reports were reviewed, verified during the walkdown, and discussed with the system engineer. Documents reviewed are listed in the attachment.

.2 Partial System Walkdowns (71111.04Q - Three Samples)

The inspectors performed the following partial system walkdowns:

C On October 12 to October 17, 2006, the B Primary Component Cooling Water System (PCCW) during the refueling outage and while the A PCCW was out-of-service; C On October 24, 2006, the B cooling tower during work on the A cooling tower pump and ocean intake; and C On October 24 and 25, 2006, the charging system following maintenance on the system.

The inspectors conducted a walkdown of each system to verify that the critical portions of the systems, such as valve positions, switches, and breakers, were correctly aligned in accordance with Seabrook's procedures and to identify any discrepancies that may have had an effect on operability. The inspectors reviewed applicable piping and instrumentation drawings and operational lineup procedures to support the walkdowns and verify proper system alignment. Documents reviewed are listed in the attachment.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

(71111.05Q - Fourteen Samples)

The inspectors examined several areas of the plant to assess: 1) the control of transient combustibles and ignition sources; 2) the operational status and material condition of the fire detection, fire suppression, and manual fire fighting equipment; 3) the material condition of the passive fire protection features (fire doors, fire dampers, fire penetration seals, etc.); and 4) the compensatory measures for out-of-service or degraded fire protection equipment. The following areas were inspected:

C Containment Building, 25' elevation; C Containment Building, 0' elevation; C Containment Building, -26' elevation; C Emergency Feedwater Pump Building, 27' elevation; C Turbine Building, all elevations; C General Area - Primary Auxiliary Building, 7' elevation;

• Demin Alley - Primary Auxiliary Building, 7' elevation; C A Charging Pump Room - Primary Auxiliary Building, 7' elevation; C Supplemental Emergency Power System Buildings; C Security Guardhouse, 8' and 21' elevations; C Security Guardhouse, 35' and 47' elevations; C Lube Oil Storage Building; C Fire Pumphouse Building; and C Transition Structures.

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

Analysis.

Documents reviews are listed in the attachment.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance (71111.07A - One Sample)

a. Inspection Scope

The inspectors examined the A PCCW heat exchanger to determine whether the heat exchanger could fulfill its design function. During the refueling outage, Seabrook conducted a tube sheet inspection of the heat exchanger and identified significant debris accumulation. This issue was documented in Condition Report (CR) 06-12389. The inspectors reviewed the engineering evaluation of the debris, EE-06-038, Evaluation of 1-CC-E-17-A, Train "A" PCCW Heat Exchanger Fouling Event, Revision 0, and evaluated the potential impact on past operability. The inspectors evaluated the results against Inspection Manual Part 9900, "Operability Determinations and Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety" and Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment.

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection (ISI) (71111.08P - Eight Samples)

a. Inspection Scope

The purpose of this inspection was to assess the effectiveness of Seabrooks ISI program for monitoring degradation of the reactor coolant system boundary, risk significant piping system boundaries, and the containment boundary. The inspectors assessed 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 observed portions of selected nondestructive examination (NDE)activities, reviewed documents, and included samples of NDE activities associated with the repair and replacement of safety-related pressure boundary components. The sample selection was based on sample availability, the inspection procedure objectives, and risk significance. Specifically, the inspectors focused on components and systems where degradation would result in a significant increase in risk of core damage. These reviews were conducted to determine that the activities were performed in accordance with ASME Code requirements. The inspectors reviewed a sample of examination reports and CRs initiated as a result of ISI examinations during this and previous outages to evaluate Seabrooks 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 control rod drive mechanism (CRDM) head penetrations, components above the reactor pressure vessel (RPV) head and bottom mounted instrumentation (BMI) penetrations in the lower head. Also, the inspectors reviewed a sample of CRs initiated as a result of the inspections performed in accordance with Seabrooks boric acid control program. The inspectors selected CRs that identified evidence of leakage which could cause degradation of safety significant components and had been identified and characterized as active or inactive coolant leakage. The inspectors reviewed operability evaluations and assessed that the corrective actions performed were consistent with the requirements of the ASME Code and 10 CFR 50, Appendix B, Criterion XVI.

The inspectors reviewed portions of the steam generator management plan, degradation assessment, and the final operational assessment to evaluate the steam generator inspection and management program. The inspectors reviewed plant specific steam generator information, tube inspection criteria, integrity assessments, degradation modes, and tube plugging criteria. The inspectors assessed the implementation of the steam generator tube inspection program by observation of selected tube inspections, interviews with data acquisition personnel, and review of a sample of the data evaluations and conclusions by analysis personnel.

Seabrook conducted eddy current testing (ECT) of one hundred percent (with the exception of two rows which were examined for partial length) of the tubes in each of the four steam generators to identify and quantify tube degradation mechanisms and to confirm tube integrity. The inspectors observed a sample of tubes examined to verify Seabrooks examination of the entire tube length. Also, the inspectors reviewed examination data for selected tubes from the C steam generator. The selected tubes failed the acceptance criteria for "wall thinning" (40 percent thru wall) and were plugged.

The inspectors reviewed the acquired data, characterization, and disposition of each of the indications to assess the problem identification and corrective action aspect of the steam generator inspection program. No tubes were identified as candidates for in-situ pressure testing during the previous and current inspection. This activity represents completion of one inspection sample.

The inspectors observed portions of five in-process NDE activities and performed a documentation review of additional examinations. The activities included volumetric and surface examinations as follows:

C Ultrasonic Test (UT), volumetric, charging system, 8 inch diameter 0.322 inch wall thickness, stainless steel SA 312 Type 304, butt weld, field weld 53; C Magnetic Particle Test (MT), surface, Feedwater (FW), 16 inch diameter 1.031 inch wall thickness, carbon steel SA 106 Grade B, butt weld, field weld 21; C Liquid Penetrant Test, surface, charging system (CS), 8 inch diameter 0.322 inch wall thickness, stainless steel SA 312 Type 304, butt weld, field weld 53; C Eddy Current Test, volumetric, reactor coolant system (RCS), Steam Generator (SG) Tubes, four tubes from SG "C", two tubes (R51C45 and R51C69) exhibiting vibration bar wear and two tubes (R59C57 and R58C54) exhibiting presence of foreign material in contact with the tubes; and C Wet Fluorescent MT, surface, Reactor vessel Head (RPV) closure studs, 7 inch diameter, alloy steel SA 540 Grade B-24, Studs #33 and 34.

This NDE inspection activity represents completion of five inspection samples.

The inspectors reviewed documentation of two samples (Work Orders 0615037 and 0522187) of a repair/replacement activity which involved welding on an ASME pressure boundary and required the development and implementation of an ASME Section XI repair/replacement plan. The activity was performed during the previous operating cycle and involved the repair of through-wall leaks in the service water system. The inspectors reviewed the ASME Section XI plan, replacement material, weld procedure specifications and qualifications, welder qualifications, weld filler metals, specified nondestructive tests, acceptance criteria, and post-work testing. This activity represents completion of one inspection sample.

The inspectors visually examined selected portions of the containment liner to determine compliance with the requirements of ASME Section XI, IWE (requirements for class MC and Metallic Liners of Class CC components). This activity represents completion of one inspection sample.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

Quarterly Resident Inspector Review (71111.11Q - One Sample)

a. Inspection Scope

The inspectors observed the conduct of licensed operators during an annual simulator examination on November 6, 2006. The inspectors reviewed the simulators physical fidelity in order to verify similarities between the Seabrook control room and the simulator. The inspectors examined the operators ability to perform actions associated with high-risk activities, the Emergency Plan, previous lessons learned items, and the correct use and implementation of procedures. The inspectors observed the training evaluators critique of the operators performance and verified that deficiencies were adequately identified, discussed, and entered into the corrective action program.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

(71111.12Q - Two Samples)

Maintenance Rule Functional Failure (MRFF) Reviews The inspectors reviewed the application of the maintenance rule (MR) for a failure of a main steam isolation valve to close (work order [WO] 0633792 and CRs 06-15262, 06-14919, 06-13609, and 06-13569) and a diode failure in the A service water pump breaker (CR 06-14933). The inspectors examined the MRFF evaluations against 10 CFR 50.65 requirements and against the guidance in Nuclear Management and Resources Council (NUMARC) 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 2. The inspectors also 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.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13 - Five

Samples)

a. Inspection Scope

The inspectors reviewed the scheduling and control of four planned maintenance activities and one emergent work troubleshooting activity in order to evaluate the 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 other equipment was properly protected.

The compensatory measures were evaluated against 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 items.

C On October 19, 2006, the inspectors reviewed the plant risk configuration while the plant was in Mode 6 with the refueling cavity full, reactor vessel internals removed and refueling with new fuel assemblies (operating mode 6H3).

Seabrook had A PCCW, A residual heat removal system (RHR), A emergency diesel generator (EDG), service water ocean pumps (SW), and an offsite line out-of-service.

C On November 21, 2006, the inspectors reviewed the plant risk configuration during planned surveillance testing on the B emergency feedwater pump, several surveillances identified as potential reactor trip initiators, and emergent maintenance on the A service water pump.

C On November 29, 2006, the inspectors reviewed the plant risk configuration during planned surveillance testing on the turbine-driven emergency feedwater pump, A emergency diesel generator, and an electric fire pump.

C On December 4, 2006, the inspectors reviewed the plant risk configuration during planned surveillance testing on the B residual heat removal system, the B emergency diesel generator, and selected switchyard breakers.

C On December 19, 2006, the inspectors reviewed the troubleshooting efforts in response to the generator stator cooling pump A failure to shutdown properly due to a misaligned breaker trip switch. The inspectors reviewed the plan and attended troubleshooting meetings. The inspectors also interviewed the operators and engineers involved.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15 - Five Samples)

a. Inspection Scope

The inspectors reviewed operability evaluations and/or condition reports in order to verify that the identified conditions did not adversely affect safety system operability or 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 the Technical Specifications (TS) limiting condition for operation implications were properly addressed. The inspectors performed field walkdowns, interviewed personnel, and reviewed the following items:

C CR 06-12685, which evaluated the babbitt material wear on the thrust bearings for the start-up feedwater pump (1-FW-P-113). Seabrook determined that the wear was within bearing tolerance, but adjusted the thrust bearing set-up to optimize bearing wear. The inspectors interviewed the system engineer.

C CR 06-14479, which evaluated the increase in the loss rate from the A PCCW head tank. The inspectors reviewed Engineering Evaluation 93-500, Affect of a Seismic Event on PCCW Heat Removal Capability, Revision 0, examined the operator abnormal operating procedure, OS1212.01, PCCW System Malfunction, Revision 10, and interviewed system engineers.

C CR 06-11038, which evaluated abnormal indication during startup feedwater pump surveillance testing. The abnormal indications were caused by air in the suction and discharge piping. The inspectors reviewed the apparent cause, interviewed system engineers and operators, examined ultrasonic testing results, and evaluated the issue for past operability.

• CR 06-11049, which evaluated the diode failure light that came on during an instrumented fast-start of the A emergency diesel generator. Seabrook determined the cause of the diode failure light was a degraded contact in the "C" phase of the K1 contactor. The inspectors reviewed the troubleshooting evaluation, discussed the issue with engineering personnel, and reviewed the corrective actions.

C CR 06-13937, which evaluated the impact of the isolation of a main steam drain path on the operability of the turbine-driven emergency feedwater pump. The inspectors reviewed the engineering evaluation for past operability, completed a walkdown of the steam supply to the pump, and interviewed the design engineer.

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications

Resident Periodic Inspection (71111.17A - One Sample)

a. Inspection Scope

The inspectors reviewed a design change (05DCR003) to automatically close the letdown isolation valve (CS-V-145) when either of the containment isolation valves, CS-V-149 and CS-V-150, in the same line closed. The purpose of the modification was to eliminate an operator workaround. The workaround occurred when the isolation valves closed on a containment isolation signal, causing an upstream relief valve to lift until the operators manually closed CS-V-145. The inspectors reviewed the design change package, including the planned post-modification testing, and conducted a walkdown of the completed modification. The modification was evaluated against Seabrook procedures, the UFSAR, and NRC requirements.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19 - Six Samples)

a. Inspection Scope

The inspectors reviewed and observed post-maintenance testing (PMT) activities to ensure: 1) the PMT was appropriate for the scope of the maintenance work completed and in accordance with MA 3.5, "Post Maintenance Testing; 2) the acceptance criteria were clear and demonstrated operability of the component; and 3) the PMT was performed in accordance with procedures. The following PMT activities were reviewed:

C On October 3, 2006, WO 0632039 following replacement of the existing K1 contractor on A emergency diesel generator. The inspectors interviewed the maintenance technicians and engineers, and reviewed the work package associated with WO 0632039.

C On October 10, 2006, OX1456.92, Centrifugal Charging Comprehensive Pump Test, Revision 0, following replacement of B Charging Pump Shaft (WO 0515812).

C On October 22, 2006, WO 0605838 following inspection and maintenance on the SW pipe lining. The inspectors interviewed the maintenance technicians and the mechanical maintenance manager.

C On October 26, 2006, WO 0543613 following modifications to 1-CS-V-145 auto-closure circuitry (see Section 1R17 for additional details).

C On October 30, 2006, LS0569.02, Corrective Maintenance of Limitorque Valve Actuator Type SMB-00, Revision 1, following replacement of A Primary Component Cooling Water Isolation from A RHR heat exchanger, 1-CC-V-145 Valve Actuator (WO 0225974).

C On October 31, 2006, OS1436.19, 18 Month Anticipated Transient Without Scram (ATWS) Mitigation System Auto Actuation Surveillance, Revision 2, following replacement of a power supply in the ATWS mitigation system under WO 0520699.

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities (71111.20 - One Sample)

.1 Refueling Outage Activities

a. Inspection Scope

The inspectors reviewed operational, maintenance, and scheduling activities prior to and during the eleventh refueling outage (OR11) to evaluate Seabrooks ability to assess and manage the outage risk. Prior to the outage, the inspectors reviewed the outage plan and the risk assessment of the schedule. During the outage, the inspectors examined the following activities: shutdown of the plant; cooldown; drain down to the reactor vessel flange and mid-loop conditions; fuel handling operations; heatup; and ascension to full power operations. The inspectors reviewed applicable procedures, observed control room activities, conducted walkdowns, and interviewed key personnel.

The inspectors also conducted periodic outage reviews of the following items: clearance activities; reactor coolant system instrumentation; electrical power configuration; residual heat removal system operation; spent fuel pool cooling system operation; inventory control measures; reactivity control measures; and containment closure requirements.

Specific documents reviewed during the inspection are listed in the attachment. The inspectors evaluated the activities against Technical Specifications requirements, Seabrooks procedures, and other applicable requirements.

b. Findings

No findings of significance were identified.

.2 Inadequate Controls of a Heavy Load Lift over the Reactor Vessel

a. Inspection Scope

In addition to the inspection scope described in Section 1R20.1, the inspectors reviewed Seabrooks heavy load lifts. The inspectors reviewed Seabrook actions following identification of the load lift of a reactor coolant pump floor plug outside of the required path. The inspectors interviewed maintenance personnel and operators, examined the associated condition report, and conducted walkdowns of the safe load path. The inspectors evaluated the action against NUREG-0612, Control of Heavy Loads and Seabrook heavy load procedures.

b. Findings

Introduction.

The inspectors identified that Seabrook failed to adequately establish and implement procedural controls for a heavy load lift, which resulted in a heavy load passing over an open, partially-fueled reactor vessel. This finding was determined to be of very low safety significance (Green) and was characterized as a violation of Technical Specifications 6.7.1.a, Procedures and Programs.

Description.

On October 11, 2006, during refueling outage OR11, the inspectors observed a reactor coolant pump (RCP) floor plug being moved from its original location to a temporary laydown area in primary containment. The inspectors observed the heavy load lift to verify that if the load were to drop, it could not impact the fuel in the core. This is required to ensure that the plant can maintain safe shutdown conditions or permit continued decay heat removal. NUREG-0612, Control of Heavy Loads, states that if the fuel in the core were damaged in the event of a heavy load being dropped on it and the fuel is highly radioactive due to its irradiation history, the potential releases of radioactive material could result in offsite doses that exceed 10 CFR Part 100 limits.

The RCP floor plug was considered a heavy load and was lifted using the containment polar crane. Seabrooks North American Lifting and Rigging Standard (NALS),

Guidelines NUREG-0612 Lifts with Mobile and Fixed Cranes, provides guidance for control of heavy loads at nuclear power plants. Based on the guidance, Seabrook developed safe load path drawing, 1-NHY-805277, to determine the acceptable path the RCP floor plug could travel. The inspectors observed that a portion of the RCP floor plug passed over the open reactor vessel containing irradiated fuel as the load was moved past the refueling cavity. The inspectors questioned licensee personnel regarding the heavy load path. Based on the inspectors observations, licensee personnel initiated condition report, CR 06-12273, to evaluate the issue. A stand-down of the affected rigging personnel immediately occurred and subsequent moves were conducted within the safe load path. Seabrooks root cause analysis determined that the Stations controls for the heavy load lift were inadequate. The inspectors determined that this issue was a performance deficiency because Seabrook had failed to establish and implement procedural controls to move the RCP floor plug within a safe load path inside containment.

Analysis.

This finding impacted the Barrier Integrity Cornerstone. The finding was more than minor because it could be reasonably viewed as a precursor to a significant event because a portion of the heavy load traveled over the reactor vessel that contained irradiated fuel and the reactor vessel head was removed for refueling activities. If the heavy load was dropped on the opened, partially-fueled reactor vessel, damage to the fuel assemblies and subsequent release to the atmosphere with the containment hatch removed could have occurred. Traditional enforcement did not apply since the issue did not have any actual safety consequences; did not impact the NRCs ability to perform its regulatory function; and did not involve any willful aspects. This finding was not suitable for a significance determination process (SDP) evaluation, but has been reviewed by NRC management, in accordance with IMC 0612, Section 05.04.c, and was determined to be a finding of very low safety significance (Green). No specific SDP appendix relates to fuel damage during shutdown due to heavy load drops. The inspectors consulted with the regional Senior Reactor Analyst and determined that the finding was not greater than very low safety significance (Green) because:

(1) the polar crane was in good working condition and had no known deficiencies that would have jeopardized the cranes ability to lift the load;

(2) the duration of the heavy load lift over the reactor cavity and vessel was short; and

(3) based on NUREG-1774, the estimated likelihood of dropping the load was very low. This finding has a cross-cutting aspect in the area of human performance (sub-category resources) because Seabrook did not have complete and accurate procedural controls to assure safety of the heavy load lift.

Enforcement.

Technical Specification 6.7.1.a, Procedures and Programs, requires that written procedures be established and implemented covering the activities in Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, Appendix A. Regulatory Guide 1.33, Section 9.0, Procedures for Performing Maintenance, requires, in part, that maintenance that can affect the performance of safety-related equipment be properly pre-planned and performed in accordance with written procedures or drawings appropriate to the circumstances. Contrary, to the above, on October 11, 2006, Seabrook failed to properly pre-plan and perform the heavy load lift as described in drawing, 1-NHY-805277, which defines the safe load paths in the vicinity of the reactor vessel. As a result, the heavy load lift of the RCP floor plug was moved outside the boundaries of a safe load path and passed over the open reactor vessel which contained irradiated fuel, a safety-related component. Because this finding was of very low safety significance and Seabrook entered this finding into their corrective action program (CR 06-12273), this finding is being treated as an NCV consistent with Section VI.A.1 of the Enforcement Policy (NCV 05000443/2006005-01, Inadequate Controls of a Heavy Load Lift over the Reactor Vessel).

.3 Inadequate Procedural Compliance Results in Inadvertent Dilution during Shutdown

a. Inspection Scope

In addition to the inspection scope described in Section 1R20.1, the inspectors reviewed reactivity manipulations during the refueling outage. Specifically, the inspectors reviewed the details associated with an inadvertent dilution of the RCS. The inspectors interviewed operators and operation managers, reviewed the condition report and root cause analysis, and observed the post-job critique. The inspectors evaluated Seabrook actions against Technical Specification requirements and Seabrook procedural requirements.

b. Findings

Introduction.

During shutdown operations, operators inadvertently performed a 500-gallon dilution instead of a planned blended makeup to the reactor coolant system.

This occurred due to operators not properly implementing the procedure for realignment of the boric acid system. This self-revealing finding was determined to be of very low safety significance (Green) and was characterized as a violation of Technical Specification 6.7.1.a, Procedures and Programs.

Description.

On October 25, 2006, operators inadvertently performed a 500-gallon dilution of the reactor coolant system. At the end of the dilution, operators recognized that the planned blended makeup (boric acid and demineralized water) had only been demineralized water. Operators added boric acid as an immediate action and restored the reactor coolant system to the desired boron concentration. The actual boron concentration of the reactor coolant system was reduced from 2588 parts per million (ppm) to 2548 ppm as a result of the unexpected dilution. Seabrook remained above the required Technical Specification shutdown margin limit of 1875 ppm.

Seabrook determined the root cause of this event was a loss of configuration control of the boric acid storage system due to a lack of procedure use in accordance with established standards. Seabrook identified multiple instances of inadequate procedure usage including the alignment of the boric acid storage system following batching to the A boric acid tank. This resulted in isolating the normal flow path from the boric acid storage system for boric acid additions and blended makeups. Seabrook identified the following contributing causes: 1) ineffective supervisory oversight by the shift manager and the unit supervisor; 2) inadequate monitoring of blended makeup as required by the procedure; 3) workload priorities not set by the shift manager and the unit supervisor; 4)inadequate peer checks by multiple individuals; and 5) inappropriate alarm response during the blended makeup. Seabrook developed numerous corrective actions to address the root and contributing causes with the focus on improving the station adherence to standards.

Seabrooks inadequate adherence to procedures was a performance deficiency.

Seabrook had multiple procedure adherence issues and several opportunities to identify and correct the issue prior to the inadvertent dilution. This event was considered self-revealing due to the clear control room indications following conclusion of the 500-gallon dilution.

Analysis.

This finding impacted the Mitigating Systems Cornerstone. The finding is more than minor because if left uncorrected it would become a more significant safety concern. Specifically, if the dilution occurred while the plant was on-line, this would have resulted in a more significant reactivity change and potential overpower condition. Using Appendix G, "Shutdown Operations," of IMC 0609, Significance Determination Process," dated November 22, 2005, the finding was determined to be of very low safety significance (Green) since this reactivity change did not result in exceeding the Technical Specification shutdown margin requirements (Reference Checklist 4 in Appendix G). This finding has a cross-cutting aspect in the area of human performance (sub-category work practices) because personnel did not follow procedural compliance standards.

Enforcement.

Technical Specification 6.7.1.a, Procedures and Programs, requires that written procedures be implemented covering the activities in Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, Appendix A. Regulatory Guide 1.33 requires procedures for operation of the chemical and volume control system.

Seabrook procedures OS1008.02, Boric Acid Preparation and Transfer to the Storage Tanks, Revision 10, and OS1008.03, Boric Acid Tanks Recirculation and Sampling, Revision 9, provide direction for the proper realignment of the boric acid storage system (a portion of the chemical and volume control system) following a batching evolution on a tank.

Contrary to the above, on October 25, 2006, Seabrook operators did not properly implement the procedure for realignment of the boric acid system. This resulted in improper alignment of the system and subsequent inadvertent dilution of the reactor coolant system during a blended makeup later on October 25. Because this violation was of very low safety significance and Seabrook entered this finding into its corrective action program (CR 06-13475), this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 05000443/2006005-02, Inadequate Procedural Compliance Results in Inadvertent Dilution during Shutdown).

1R22 Surveillance Testing (71111.22 - Seven Samples)

a. Inspection Scope

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 Technical Specifications and surveillance procedures.

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

C On October 5, 2006, OS1001.10, Reactor Coolant System Draining Using Nitrogen Injection to the Steam Generators, Revision 0, and OS1000.12, Operation with RCS at Reduced Inventory/Midloop Conditions, Revision 4.

C On October 5, 2006, E1803.003, Reactor Containment Type B and C Leakage Rate Tests, Revision 6 (Penetrations X-16 and X-18).

C From October 9 through 20, 2006, RS06-01-03, Ultrasonic Fuel Assembly Cleaning Site Procedure for Seabrook, Revision 0.

C On October 26, 2006, OX1426.21, Diesel Generator 1B 18 Month Operability and Engineered Safeguards Pump and Valve Response Time Testing Surveillance, Revision 3.

C On October 26, 2006, OX1456.90, Auto SI [safety injection], Phase A, Phase B, CBS [Containment Building Spray], CVI [Containment Vessel Isolation] & CBA

[Control Building Air Handling] Actuation and Manual SI, Phase A, Phase B, CBS

& CVI Actuation 18 Month Surveillance - Train B, Revision 1.

C On November 1, 2006, OX1436.13, Turbine Driven Emergency Feedwater Pump Post Cold Shutdown or Post Maintenance Surveillance and Comprehensive Pump Test, Revision 8.

C On November 3, 2006, RS1737, Post Refueling Low Power Physics Testing, Revision 4.

b. Findings

No findings of significance were identified.

EMERGENCY PREPAREDNESS 1EP6 Drill Evaluation (71114.06 - One Sample)

a. Inspection Scope

The inspectors reviewed the operators emergency classification and notification completed during requalification training on November 6, 2006 (See Section 1R11).

The inspectors evaluated the results against Seabrooks Emergency Response Manual 1.1, Classification of Emergencies, Revision 42 and Nuclear Energy Institute (NEI)99-02, Regulatory Assessment Indicator Guideline, Revision 4.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access to Radiologically Significant Areas (71121.01 - Ten Samples)

a. Inspection Scope

During the period October 23 to 26, 2006, the inspectors conducted the following activities to verify that Seabrook was properly implementing physical, administrative, and engineering controls for access to locked high radiation areas, and other radiologically controlled areas (RCA) during the refueling outage (OR11), and 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 Seabrooks procedures.

This activity represents the completion of ten samples relative to this inspection area and with the eleven samples previously completed in January 2006 (documented in inspection report 05000443/2006002), the annual inspection requirement of twenty-one samples has been accomplished.

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 removal of nozzle dams from the steam generators and decontamination of the drive shaft storage rack (DSSR), located in the reactor cavity.

Tasks in the PAB included sluicing spent resin and demobilization from outage valve maintenance testing and repairs. 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, Fuel Storage Building, and Waste Processing Building with the sites Radiation Protection Manager. The inspectors 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 to determine if ED placement was in the highest dose field. The inspectors also evaluated dose/dose rate alarm set points to determine if the set points were consistent with the area radiological conditions and plant policy. The inspectors verified that the workers were knowledgeable of the actions to be taken when the electronic dosimeter alarms or malfunctions. Work activities reviewed included spent resin sluicing from the primary resin bed (RWP 06-051), steam generator nozzle dam removal (RWP 06-037), and DSSR decontamination/survey (RWP 06-06-053).

The inspectors reviewed the radiological controls applied to recently completed outage tasks to evaluate the effectiveness of controlling exposure. Included in the review were split pin replacement (RWP 06-065), reactor head inspections (RWP 06-064), and installation of debris interceptors in the containment building sump (RWP 06-061).

Problem Identification and Resolution The inspectors reviewed elements of Seabrooks corrective action program related to controlling access to radiologically controlled areas, completed since the last inspection of this area, to determine if problems were being entered into the program for resolution.

Details of this review are contained in Section 4OA2.4 of this report.

Jobs-In-Progress The inspectors observed aspects of various outage-related tasks being 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 steam generator nozzle dam removal, DSSR decontamination, and spent resin sluicing.

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 controls and procedures.

These special areas included under the reactor vessel spent fuel transfer routes in containment, spent resin sluicing paths and spent resin storage locations in the primary auxiliary building, and irradiated hardware stored in the spent fuel pool. The inspectors evaluated the prerequisite 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 licensee procedures did not substantially reduce the effectiveness and level of worker protection.

Keys to locked high radiation areas (LHRA) and very high radiation areas that are maintained at the radiation protection control point and in the control room, 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 briefings, observing jobs-in-progress, and questioning individuals regarding their knowledge of radiological controls and contamination control measures 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

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02 - Eight Samples)

a. Inspection Scope

During the period October 23 to 26, 2006, the inspectors conducted the following activities to verify that Seabrook was properly implementing operational, engineering, and administrative controls to maintain personnel exposure as low as is reasonably achievable (ALARA) for tasks conducted during OR11. Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, applicable industry standards, and Seabrooks procedures.

This inspection activity represents completion of eight samples relative to this inspection area; partially completing the biennial inspection requirement of fifteen samples.

Radiological Work Planning The inspectors reviewed pertinent information regarding the sites cumulative exposure history, current exposure trends, and ongoing activities to assess current performance and exposure challenges The inspectors reviewed the refueling outage work scheduled during the inspection period and the associated work activity exposure estimates. Scheduled work included steam generator nozzle dam removal, DSSR decontamination, spent resin sluicing, and containment de-mobilization. As part of this review, the inspectors evaluated the dose estimates for these jobs and reviewed the basis for making the exposure forecasts.

The inspectors reviewed the procedures associated with maintaining worker dose ALARA and with estimating and tracking work activity specific exposures. The inspectors reviewed the OR11 Project Dose Summary Report, detailing the worker estimated and actual exposures, through October 25, 2006, 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 an estimated dose for tasks associated with these work activities. Jobs reviewed included reactor vessel disassembly/re-assembly (AR 06-01), steam generator primary/secondary maintenance (AR 06-02/03), in-service inspection (AR 06-04), cavity decontamination (AR 06-05),valve maintenance (AR 06-07), scaffolding installation/removal (AR 06-11), split pin replacement project (AR 06-13), reactor head inspections (AR 06-14), and sump debris interceptor installation (AR 06-15).

The inspectors evaluated the departmental interfaces between radiation protection, operations, maintenance, 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 (Nos. 06-01 through 06-04), reviewing outage-related Nuclear Assurance Quality Reports, observing jobs-in-progress, and attending the pre-job briefing for spent resin sluicing.

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 the split pin replacement project, scaffolding installation, debris interceptor sump modification, and steam generator eddy current testing.

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 conducting this evaluation, the inspectors reviewed temporary shielding requests (selected TSR Nos. 06-TSR-01 through 06-TSR-33), split pin replacement project prerequisite, shutdown chemistry requirements, and steam generator eddy current testing preparations.

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 Seabrooks method for adjusting exposure estimates, and re-planning work, based on work progress. This review included evaluating the basis for the Radiation Safety Committee to lower the outage stretch goal from 84 person-rem to 76 person-rem, during the outage, due to successes realized in minimizing exposure for various outage projects.

The inspectors reviewed Seabrooks 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 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 being performed for steam generator nozzle removal, reactor cavity decontamination, containment building demobilization, and spent resin sluicing to verify that prerequisite radiological controls were implemented and that workers were knowledgeable of work area radiological conditions and implementing ALARA practices.

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

Source Term Reduction Control The inspectors reviewed the current status and historical trends of the sites source terms. Through interviews with the Chemistry Supervisor and Radiation Protection Manager, the inspectors evaluated the effectiveness of Seabrooks source term control strategy. Specific strategies being employed by Seabrook included post-shutdown peroxide flushes of the reactor coolant system, use of a macro porous resin for coolant cleanup, use of a submersible demineralizer for reactor cavity cleanup, relocating irradiated components away from work areas, and customizing temporary shielding for the reactor head.

Radiation Worker Performance The inspectors observed radiation worker and health physics technician performance during steam generator removal and DSSR decontamination 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 an exposure significant task, spent resin sluicing, and observed the donning of Delta Suits and multi-dosimetry of workers performing steam generator nozzle dam removal to evaluate worker proficiency.

The inspectors reviewed condition reports, related to radiation worker and radiation protection technician errors, and personnel contamination reports 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 - Two Samples)

.1 Occupational Exposure Control Effectiveness

a. Inspection Scope

The inspectors reviewed implementation of Seabrooks Occupational Exposure Control Effectiveness Performance Indicator Program. Specifically, the inspectors reviewed condition reports, and associated documents, for occurrences involving locked high radiation areas, very high radiation areas, and unplanned exposures against the criteria specified in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 4, to verify that all occurrences that met the NEI criteria were identified and reported as performance indicators. This inspection activity represents the completion of one sample relative to this inspection area; completing the annual inspection requirement.

b. Findings

No findings of significance were identified.

.2 RETS/ODCM Radiological Effluent Occurrences

a. Inspection Scope

The inspectors reviewed relevant effluent release reports for the period October 1, 2005 through September 30, 2006, for issues related to the public radiation safety performance indicator, which measures radiological effluent release occurrences that exceed 1.5 mrem/quarter whole body or 5.0 mrem/quarter organ dose for liquid effluents; 5 mrads/quarter gamma air dose, 10 mrad/quarter beta air dose, and 7.5 mrads/quarter for organ dose for gaseous effluents. This inspection activity represents the completion of one sample relative to this inspection area; completing the annual inspection requirement.

The inspectors reviewed the following documents to ensure Seabrook met all requirements of the performance indicator from the fourth quarter 2005 through the third quarter 2006:

C Monthly projected dose assessment results due to radioactive liquid and gaseous effluent releases; C Quarterly projected dose assessment results due to radioactive liquid and gaseous effluent releases; and C Dose assessment procedures.

b. Findings

No findings of significance were identified.

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

.1 Routine Condition Report Screening

a. Inspection Scope

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

b. Findings

No findings of significance were identified.

.2 Annual Sample Review (71152 - One Sample)

(Open) Unresolved Item (URI) 50-443/2006-005-03, Testing of the Alternate Supply of Water to the Primary Component Cooling System

a. Inspection Scope

The inspectors followed up on a condition report (CR 06-12990) initiated during the recent refueling outage related to the inability to establish alternate cooling to the lube oil coolers for the centrifugal charging pumps (CCPs). Specifically, the CR noted that alternate cooling from the demineralized water (DM) system could not be established to the A CCP until mechanical agitation was applied. Subsequently, the DM supply to the B CCP was tested as part of the extent-of-condition review; the CR noted that tapping of the check valve was required to establish flow. The inspectors reviewed CRs, the UFSAR, procedures, and system prints, interviewed operations and engineering personnel, and walked down the applicable portions of the systems.

b. Inspection Findings The primary component cooling water (PCCW) system is the normal cooling supply for safety-related components, including the CCP lube oil coolers. The Seabrook Updated Final Safety Analysis Report (UFSAR), Section 9.2.2, states that for increased reliability the DM system and the fire protection (FP) systems can be cross-connected to the PCCW system to make up for leaks from the PCCW system and to specifically provide cooling to the CCP lube oil coolers. In the event of a seismic event, the cross-connect is backed up by the seismic Category I service water (SW) system, using the seismic portion of the FP system. The inspectors reviewed CR 06-12990 and noted that the review for the extent-of-condition was limited to testing of the supply from the DM system for alternate cooling to the B CCP lube oil cooler. The inspectors questioned operations and engineering personnel as to why the other sources of alternate cooling (the FP and SW systems) had not been tested or evaluated to determine if they would function. In addition, the inspectors questioned the basis for deciding to not perform a causal analysis to determine why both trains of alternate cooling from the DM system had failed to function. Seabrook planned to conduct performance tests to verify that the flow paths were functional, and to perform an evaluation of the failures of the DM supply. This issue remains unresolved until Seabrook completes the flow tests and the causal evaluation (URI 05000443/2006005-03, Testing of the Alternate Supply of Water to the Primary Component Cooling System). The inspectors will review the final results and determine if any performance deficiencies existed, and the potential safety significance of those deficiencies.

.3 Inservice Inspection

a. Inspection Scope

The inspectors reviewed a sample of condition reports related to inservice inspection listed in the attachment (Section 1R08), which identified indications and other nonconforming conditions discovered during this and the previous outage.

b. Findings

No findings of significance were identified.

.4 ALARA Planning and Controls

a. Inspection Scope

The inspectors reviewed Nuclear Assurance Quality Audit Report (SBK-06-02), Daily Quality Summaries for the period September 1 to October 21, 2006, recent Radiation Safety Committee meeting minutes, and 23 CRs relating to the implementation of physical, engineering, and administrative controls for performing work in radiologically controlled areas. This review evaluated the threshold for identifying problems and the promptness and effectiveness of the subsequent evaluations and corrective actions.

The review was conducted against the criteria contained in 10 CFR 20, Technical Specifications, and Seabrooks procedures.

b. Findings

No findings of significance were identified.

.5 Semi-Annual Problem Identification and Resolution Trend Review (One Sample)

a. Inspection Scope

As required by Inspection Procedure 71152, Problem Identification and Resolution, the inspectors performed a review of the Seabrook corrective action program 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 inspector corrective action program item screening. The inspectors compared and contrasted their results with the results contained in the Seabrook 2nd Quarter Corrective Action Program Quarterly Trend Report.

b. Inspection Findings No findings of significance were identified. The inspectors did not identify any appreciable trends that the licensee had not already identified. The 2nd Quarter Trend report noted that work practices continues to be the major contributor for all levels of CRs.

4OA5 Other Activities

.1 Reactor Pressure Vessel Head And Vessel Head Penetration Nozzles Inspection

a. Inspection Scope

The inspectors reviewed Seabrooks examination activities performed in response to NRCs First Revised Order Modifying Licenses, "Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors,"

EA-03-009, dated February 20, 2004. In addition, the inspectors reviewed the Seabrooks Relaxation Request of the First Revised Order (March 30, 2006). Seabrook requested relaxation from the order as it pertains to the minimum inspection coverage requirement due to design limitations affecting inspectability in full compliance with the order. The NRC staff had previously authorized the alternative inspection for specified RPV penetration nozzles at Seabrook for the life of the Order.

The inspector reviewed the licensees inspection methods to detect evidence of leakage and/or cracking of RPV head penetrations (control rod drive mechanism and vessel head vent). The inspectors also reviewed the examination procedures to determine whether they provided detailed process guidance and acceptance criteria for performance of the examinations. Seabrook performed a bare metal visual examination of the top of the RPV to include all of the CRDM penetrations and the RPV head vent to evaluate the integrity of vessel head and penetration intersections. This examination of one hundred percent of the head penetrations was to confirm the absence of flaws and boric acid deposits. The inspectors observed the bare metal visual examination of penetration nozzles #56, 44, 33 and 19 and noted the absence of boric acid deposits at the head to penetration intersection. The inspectors also observed the performance of a visual examination for indications of boric acid leaks from pressure-retaining components above the RPV head.

The inspectors evaluated the effectiveness of the ECT and UT test methods to detect surface and subsurface flaws in the CRDM penetration. Also, the combined process of ECT and UT was used to confirm the absence of a "leak path" from any failure in the vicinity of the vessel head penetration to head in the "shrink fit" location. The inspectors reviewed the UT reports, ECT reports, and visual data displays for penetrations #17, 20, 41, 46 and 53. The examination procedures required that anomalies, deficiencies, and discrepancies identified are documented and evaluated in accordance with Seabrooks corrective action program.

The inspectors interviewed examination personnel, data analysts and engineering personnel, and reviewed training and qualification records to determine that licensee personnel were properly trained to perform the bare metal visual inspection of the reactor vessel head and the ECT and UT of the head penetrations.

b. Findings

No findings of significance were identified.

.2 Administrative Issues

The inspectors previously completed an inspection that was documented in NRC Inspection Report 05000443/2006010, Section 4OA5. The inspection was performed as a problem identification and resolution annual sample (71152), but was not documented as a completed sample due to an administrative error. This documents that a problem identification and resolution annual sample was completed.

In NRC Inspection Report 05000443/2006004, Section 4OA5.1, the inspectors documented their review of mitigating systems performance indexes for the emergency alternate current power, high pressure safety injection, auxiliary feedwater, residual heat removal, and cooling water support systems. The inspection was documented as completion of Temporary Instruction 2515/169, Mitigating Systems Performance Index Verification, but was not documented as a completion of the different system performance indicator reviews. This documents that the annual samples for the four system performance indicators were completed.

4OA6 Meetings, including Exit

Exit Meeting Summary

The inspectors presented the inspection results to Mr. G. St. Pierre on January 9, 2007, following the conclusion of the period. The licensee acknowledged the findings presented. The licensee did not indicate that any of the information presented at the exit meeting was proprietary.

Site Management Visit On October 17, David Lew, Division Director, Division of Reactor Projects, Region 1, toured the site and met with Mr. G. St. Pierre and other members of licensee management.

ATTACHMENTS: 1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

P. Allen Senior Health Physics Technician

M. Bianco Supervisor, Radiological Waste Services

T. Date Senior Health Physics Technician

M. Debay Nuclear Oversight Manager

D. Perkins Health Physics Analyst

K. Douglas Work Management Supervisor

P. Freeman Engineering Director

D. Hampton Health Physics Shift Supervisor

P. Harvey Chemistry Manager

M. Kiley Station Director

M. Makowicz Plant Engineering Manager

M. OKeefe Regulatory Compliance Supervisor

D. Ritter Operations Manager

V. Robertson Senior Nuclear Analyst

M. Scannell Health Physics Shift Supervisor - Nuclear

D. Sherwin Maintenance Manager

G. St. Pierre Site Vice President

R. Sterritt Health Physics Specialist - Nuclear

R. Thurlow Radiation Protection Manager

J. Tucker Security Manager

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000443/2006005-03 URI Testing of the Alternate Supply of Water to the Primary Component Cooling System (Section 4OA2.2)

Opened and Closed

05000443/2006005-01 NCV Inadequate Controls of a Heavy Load Lift over the Reactor Vessel (Section 1R20.2)

05000443/2006005-02 NCV Inadequate Procedural Compliance Results in Inadvertent Dilution during Shutdown (Section 1R20.3)

A-1-2