IR 05000369/2008005

From kanterella
Jump to navigation Jump to search
IR 05000369-08-005, IR 05000370-08-005, IR 05000369-08-501, IR 05000370-08-501, on 9/1/2008 - 12/31/2008, McGuire Nuclear Station, Units 1 and 2, Maintenance Risk Assessments and Emergent Work Evaluation, Other
ML090290315
Person / Time
Site: Mcguire, McGuire  Duke Energy icon.png
Issue date: 01/29/2009
From: Bartley J
NRC/RGN-II/DRP/RPB1
To: Brandi Hamilton
Duke Energy Carolinas, Duke Power Co
References
EA-08-268 IR-08-005, IR-08-501
Download: ML090290315 (41)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION II

ary 29, 2009

SUBJECT:

MCGUIRE NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000369/2008005, 05000370/2008005, 05000369/2008501, AND 05000370/2008501

Dear Mr. Hamilton:

On December 31, 2008, the US Nuclear Regulatory Commission (NRC) completed an inspection at your McGuire Nuclear Station. The enclosed report documents the inspection findings which were discussed on January 13, 2009, with you and members of your staff.

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

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

This report documents one finding of very low safety significance (Green) which was determined to be a violation of NRC requirements and one finding categorized as a Severity Level IV violation under traditional enforcement. However, because of the very low safety significance and categorization at Severity Level IV, and because they were entered into your corrective action program, the NRC is treating these NRC-identified findings as non-cited violations (NCVs)

consistent with Section VI.A of the NRC Enforcement Policy. If you contest any of these NCVs, you should provide a written 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 II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the McGuire facility In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document

DPC 2 Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Jonathan H. Bartley, Chief, Reactor Projects Branch 1 Division of Reactor Projects Docket Nos. 50-369, 50-370 License Nos. NPF-9, NPF-17

Enclosure:

NRC Integrated Inspection Report 05000369/2008005, 05000370/2008005, 05000369/2008501, and 05000370/2008501 w/Attachments: (1) Supplemental Information, (2) OI Synopsis

REGION II==

Docket Nos: 50-369, 50-370 License Nos: NPF-9, NPF-17 Report Nos: 05000369/2008005, 05000370/2008005, 05000369/2008501, 05000370/2008501 Licensee: Duke Power Company, LLC Facility: McGuire Nuclear Station, Units 1 and 2 Location: 12700 Hagers Ferry Road Huntersville, NC 28078 Dates: October 1, 2008 through December 31, 2008 Inspectors: J. Brady, Senior Resident Inspector R. Eul, Resident Inspector R. Chou, Reactor Inspector (Section 1R08)

J. Fuller, Senior Reactor/Construction Inspector (Section 1R08)

H. Gepford, Senior Health Physicist (Section 4OA5.2)

L. Miller, Senior Emergency Preparedness Inspector (Sections 1EP 2, 1EP3, 1EP4, 1EP5, 4OA1.2, 4OA5.3)

D. Harmon, Nuclear Safety Professional Development Program (NSPDP) Engineer Approved by: Jonathan Bartley, Chief Reactor Projects Branch 1 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS IR05000369/2008005, IR05000370/2008005, IR05000369/2008501, IR05000370/2008501; 9/1/2008 - 12/31/2008; McGuire Nuclear Station, Units 1 and 2; Maintenance Risk Assessments and Emergent Work Evaluation, Other The report covered a three month period of inspection by two resident inspectors and an announced inspection by four region based inspectors and one NSPDP engineer. One Green non-cited violation (NCV) and one Severity Level (SL) IV NCV were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using 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 (ROP), Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealing Findings Cornerstone: Initiating Events

  • Green. The inspectors identified a NCV of 10CFR50.65(a)(4) for failure to perform an adequate risk assessment for Unit 1 when the performance of switchyard activities affected both units, and were categorized as risk significant for Unit 2. This finding was documented in the licensees corrective action program as Problem Investigation Process report (PIP) M-08-6297. No immediate corrective action was taken because the work was almost completed by the time the licensee confirmed their error. Long term corrective actions include training personnel on performing shutdown risk assessments.

This finding is greater than minor because the Unit 1 risk assessment failed to consider maintenance activities that were occurring in the switchyard that affected both units and would have resulted in a higher risk category if properly assessed and could increase the likelihood of initiating events such as loss of offsite power. The finding was determined to be of very low safety significance because the time to boil in the spent fuel pool was slightly over 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, which would have allowed sufficient time such that upon a loss of offsite power there would have been a reasonable likelihood for success of actions taken to recover off-site power. This finding has a cross-cutting aspect of decision making in the area of human performance

H.1.a]. (Section 1R13)

Cornerstone: Occupational Radiation Safety

  • SL IV. The inspectors identified a NCV of Technical Specification (TS) 5.7.2 for the licensees failure to control access to a locked-high radiation area (LHRA). Specifically, on September 30, 2006, a contract radiation protection technician (RPT) left the reactor head inspection stand LHRA barrier unlocked and unguarded from approximately 5:05 to 5:21 a.m. Dose rates as high as 10 rad/hr at 30 cm and 4 rad/hr general area were present inside the reactor head stand LHRA.

Enclosure

The significance of the violation was assessed using traditional enforcement because it involved willfulness [EA-08-268]. The safety significance of this violation was determined to be SL IV because the finding did not involve a situation with a substantial potential for exposure in excess of applicable limits and was a matter with more than a minor safety, health, or environmental significance. Although this violation involved willfulness, it was dispositioned as an NCV in accordance with Section IV.A.1 of the Enforcement Policy because the licensee identified the violation and promptly discussed it with regional health physics inspectors, the violation involved the acts of a low-level individual, the violation appears to be the isolated action of the employee without management involvement, and significant remedial action commensurate with the circumstances was taken by the licensee.

The finding was documented in the licensees corrective action program as PIP M-06-4479.

(Section 4OA5.2)

B. Licensee Identified Violations None Enclosure

Report Details Summary of Plant Status:

Unit 1 began the inspection period in the end-of-cycle (EOC) 19 refueling outage, with refueling in progress. The reactor achieved criticality on October 31, 2008, but was shutdown hours later to perform emergent maintenance on the control rod system. The reactor was subsequently taken critical and went on-line on November 12, 2008. It reached 100 percent rated thermal power on November 17, 2008, and remained there for the rest of the period.

Unit 2 began the inspection period at approximately 100 percent rated thermal power and remained there for the rest of the period.

1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity 1R01 Adverse Weather Protection

.1 Severe Weather Condition (Seasonal)

a. Inspection Scope After the licensee completed preparations for seasonal cold temperature, the inspectors discussed with the licensee their Cold Weather Program and cold weather performance test. The inspectors reviewed the completed test results for PT/0/B/4700/038, Verification of Freeze Protection Equipment and Systems, dated October 28, 2008. The inspectors walked down the Auxiliary Feedwater (CA) system and the Fueling Water Storage Tank.

This equipment was selected because their safety related functions could be affected by adverse weather (freezing conditions). The inspectors reviewed documents listed in Attachment 1 of this report, observed plant conditions, and evaluated those conditions using criteria documented in procedure IP/1/B/3250/059 and IP/2/B/3250/059, Monthly Check of Freeze Protection.

b. Findings No findings of significance were identified.

Enclosure

1R04 Equipment Alignment

.1 Partial Walkdown a. Inspection Scope The inspectors performed a partial walkdown of the following systems to assess the operability of redundant or diverse trains and components when safety equipment was inoperable. The inspectors focused on 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 determined whether selected breakers, valves, and support equipment were in the correct position to support system operation. The documents reviewed during this inspection are listed in Attachment 1 of this report.

  • 2A EDG when 2B EDG was out of service for maintenance on November 25, 2008
  • 1A Safety Injection train when the 1B Safety Injection train was out of service for maintenance on December 3, 2008 b. Findings No findings of significance were identified.

.2 Complete System Walkdown a. Inspection Scope The inspectors conducted a detailed review of the standby shutdown facility (SSF) system.

To determine the correct system alignment, the inspectors reviewed procedures, drawings, and the Updated Final Safety Analysis Report (UFSAR). Items reviewed during the inspection included: (1) valves are correctly positioned, do not exhibit leakage, and are locked as required; (2) electrical power is available, (3) system components are correctly labeled, cooled, lubricated, ventilated, etc.; (4) hanger and supports are correctly installed and functional; (5) essential system support systems are functional; (6) system performance is not hindered by debris; and (7) tagging clearances are appropriate. To determine the effect of outstanding design issues on the operability of the systems the inspectors reviewed the operator workaround list, the temporary modification list, system health reports, and other outstanding items tracked by the engineering department. In addition, the inspectors reviewed outstanding maintenance work requests/work orders and deficiencies that could affect the ability of the system to perform its function. The inspectors also discussed all open issues with the licensees system engineer. The documents reviewed during this inspection are listed in Attachment 1 of this report.

Enclosure

b. Findings No findings of significance were identified.

1R05 Fire Protection

.1 Fire Protection Walkdowns a. Inspection Scope The inspectors walked down accessible portions of the plant areas listed below to determine if they were consistent with the UFSAR and the fire protection program for defense in depth features. The features assessed included the licensees control of transient combustible material and ignition sources, fire detection and suppression capabilities, fire fighting equipment, and passive fire features such as fire barriers. The inspectors also reviewed the licensees compensatory measures for fire deficiencies to determine if they were commensurate with the significance of the deficiency. The inspectors reviewed the fire plans for the areas selected to determine if they were consistent with the fire protection program and presented an adequate fire fighting strategy. The resident inspectors inspected two Fire Areas (Fire Area 2a and 3a) to ensure that the licensee had established compensatory measures for Hemyc installations in accordance with the licensees response to U.S. Nuclear Regulatory Commission Generic Letter 2006-003, Potentially Nonconforming Hemyc and MT Fire Barrier Configuration, dated April 10, 2006.

The documents reviewed during this inspection are listed in Attachment 1 of this report.

  • Standby shutdown facility (Fire Area SSF)

b. Findings No findings of significance were identified.

Enclosure

1R08 Inservice Inspection (ISI) Activities - Unit 1 (71111.08P)

.1 Non-Destructive Examination (NDE) Activities and Welding Activities a. Inspection Scope From September 29 to October 10, 2008, the inspectors reviewed the implementation of the licensees ISI program for monitoring degradation of the reactor coolant system boundary and risk significant piping boundaries. The inspectors activities consisted of an on-site review of NDE and welding activities to evaluate compliance with the applicable edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI (Code of record: 1998 Edition through 2000 Addenda), and to verify that indications and defects (if present) were appropriately evaluated and dispositioned in accordance with the requirements of the ASME Code,Section XI acceptance standards. Documents reviewed are listed in Attachment 1 of this report.

The inspectors observation and/or review of NDE activities specifically covered examination procedures, NDE reports, equipment and consumables certification records, personnel qualification records, and calibration reports (as applicable) for the following examinations:

- Film Review The inspectors also reviewed documentation for the following indications, which were accepted for continuous service:

  • Visual Examination (VT) of 1FWST-SUPPORT, Report No. VT-07-089, Disposition -

Problem Investigation Process (PIP) M-06-4832

  • VT of 1-MCR-NV-1063, Report No. VT-07-073, Disposition - Work Order (WO)

01698701 01, 1XNAISI-INSP: ISI Hangers in Reactor Building 1EOC18 Enclosure

The inspectors review of welding activities specifically covered the welding activity listed below in order to evaluate compliance with procedures and the ASME Code. The inspectors reviewed the work order, repair and replacement plan, weld data sheets, welding procedures, procedure qualification records, welder qualification records, and NDE reports.

.2 PWR Vessel Upper Head Penetration Inspection Activities a. Inspection Scope The inspectors reviewed the licensees activities related to the Bare Metal Visual examination of the reactor pressure vessel upper head nozzles and the visual examination to identify potential boric acid leaks from pressure-retaining components above the reactor pressure vessel upper head. These activities were reviewed to verify licensee compliance with the regulatory requirements of NRC Order EA-03-009 Modifying Licenses dated February 20, 2004. Documents reviewed are listed in Attachment 1 of this report.

The inspectors reviewed the licensees visual inspection results of leakage locations above the reactor pressure vessel upper head from the Unit 1 EOC 19 refueling outage.

This visual inspection included visual inspection of the following:

  • Mirror insulation at reactor vessel flange,
  • Conoseal flanges and thermocouple fittings,
  • Reactor level instrumentation system instrument tubing and isolation valve,
  • Control Rod Drive Mechanism intermediate canopy seal welds b. Findings No findings of significance were identified.

.3 Boric Acid Corrosion Control (BACC) Inspection Activities a. Inspection Scope The inspectors reviewed the licensees BACC program activities to ensure implementation with commitments made in response to NRC Generic Letter 88-05, Enclosure

Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and applicable industry guidance documents. Specifically, the inspectors performed an on-site record review of procedures and the results of the licensees containment walk-down inspections performed during the Unit 1 fall 2008 outage. The inspectors also interviewed the BACC program owner and conducted an independent walk-down of the reactor building to evaluate compliance with licensees BACC program requirements and verify that degraded or non-conforming conditions, such as boric acid leaks identified during the containment walk-down, were properly identified and corrected in accordance with the licensees BACC and corrective action programs. The inspectors also reviewed a recent self-assessment for the McGuire BACC program, dated November 13-15, 2007.

Documents reviewed are listed in Attachment 1 of this report.

The inspectors reviewed a sample of engineering evaluations completed for evidence of boric acid found on systems containing borated water to verify that the minimum design code required section thickness had been maintained for the affected components. The inspectors selected the following evaluations for review:

  • M-07-2002 - CVCS Excess Letdown Heat Exchanger 1NVHX00036
  • M-08-3147 - Safety Injection Pressure Gauge 2NIPG5240 b. Findings No findings of significance were identified.

.4 Steam Generator (SG) Tube Inservice Inspection a. Inspection Scope On October 6 - 10, 2008, the inspectors reviewed activities, plans, condition monitoring and operational assessments, the pre-outage degradation assessment, and procedures for the inspection and evaluation of the steam generator Inconel Alloy 690TT tubing for Unit 1 SGs A, B, C, and D to determine if the activities were being conducted in accordance with Technical Specifications (TS) and applicable industry standards. Data gathering, analysis, and evaluation activities were also reviewed. Documents reviewed are listed in Attachment 1 of this report.

The inspectors reviewed data results to verify the adequacy of the licensee=s primary, secondary, and resolution analyses. The inspectors also observed and reviewed the activity of the foreign object inspection and retrieval.

The inspectors reviewed equipment, data operators, and analyst certifications and qualifications, including medical exams.

Enclosure

The inspectors reviewed data for the following tubes:

SG A: R88C73, R99C76, R90C69, R111C66, and R114C67 SG B: R89C72, R97C72, R96C75, R116C75, and R94C75 SG C: R110C61, R111C60, and R108C59 SG D: R88C75, R44C133, and R28C125

[Note: The tube maximum through-wall depth wear from foreign objects during the inspection period was 27 percent at SG B tube R97C72 in two refueling cycles.]

b. Findings No findings of significance were found.

1R11 Licensed Operator Requalification a. Inspection Scope On November 13, 2008, the inspectors observed operators in the plants simulator during licensed operator requalification training to determine the effectiveness of licensed operator requalification training required by 10CFR55.59 and the adequacy of operator performance. The inspectors focused on clarity and formality of communication, use of procedures, alarm response, control board manipulations, group dynamics, and supervisory oversight. The inspectors observed the post-exercise critique to determine whether the licensee identified deficiencies and discrepancies that occurred during the simulator training. The inspectors observed the shift crews response to the scenarios listed below. The documents reviewed during this inspection are listed in Attachment 1 of this report.

  • OP-MC-SRT-05
  • OP-MC-SRT-54 b. Findings No findings of significance were identified.

1R12 Maintenance Effectiveness a. Inspection Scope The inspectors reviewed the samples listed below for items such as: (1) appropriate work practices; (2) identifying and addressing common cause failures; (3) adequacy of corrective actions; (4) scoping in accordance with 10 CFR 50.65(b) of the maintenance rule; (4) characterizing reliability issues against performance criteria; (5) trending key parameters for condition monitoring; (6) charging unavailability for performance; (7) classification and reclassification in accordance with 10 CFR 50.65(a)(1) or (a)(2);

and (8) appropriateness of performance criteria for structures, systems, and components (SSCs)/functions classified as (a)(2); and/or (9) appropriateness and adequacy of goals Enclosure

and corrective actions for SSCs/functions classified as (a)(1). For each item selected, the inspectors performed a detailed review of the problem history and surrounding circumstances, evaluated the extent of condition reviews as required, and reviewed the generic implications of the equipment and/or work practice problem. The documents reviewed during this inspection are listed in Attachment 1 of this report.

  • Unit 2 Train A EDG fuel oil transfer pump did not auto-stop as required b. Findings No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation a. Inspection Scope The inspectors reviewed the licensees risk assessments and the risk management actions used to manage risk for the plant configurations associated with the activities listed below. The inspectors assessed whether the licensee performed adequate risk assessments, and implemented appropriate risk management actions when required by 10CFR50.65(a)(4). For emergent work, the inspectors also verified that any increase in risk was promptly assessed, and that appropriate risk management actions were promptly implemented. The documents reviewed during this inspection are listed in Attachment 1 of this report.

  • Planned Unit 2 Orange risk condition for installation of 1B service water strainer during Unit 1 outage that involved work that was a risk threat to Unit 2
  • Planned Unit 1 Orange risk for cold reduced inventory/mid-loop operations to remove steam generator nozzle dams
  • Review of Unit 1 Defense in Depth (DID) on October 9 for adequacy of risk assessment during work in the switchyard
  • Planned Operational Risk Assessment Management-Orange risk for removal of 2B Nuclear Service Water (RN) strainer backwash function from service on November 7, 2008, to install valves for upgrade of strainer differential pressure indication to safety-related.

b. Findings Introduction: The inspectors identified a green NCV of 10CFR50.65(a)(4) for failure to perform an adequate risk assessment for Unit 1 when the performance of switchyard activities affected both units, and were categorized as risk significant for Unit 2.

Description: On October 9, 2008, the inspectors identified that there were switchyard activities being accomplished that the licensees risk assessment considered risk significant for Unit 2, but were not considered risk significant for Unit 1, which was shutdown in a scheduled refueling outage. During the period of the switchyard work, Enclosure

both Unit 1 safety buses were being powered from Unit 2. The licensee reviewed procedure NSD 403, Shutdown Risk Management (Modes 4, 5, 6, and No-Mode) per 10CFR50.65(a)(4), and concluded that the switchyard activities should have resulted in the Defense-in-Depth (DID) sheet not taking credit for No Safety Significant Switchyard Work In Progress. With a proper DID assessment the Power Availability Key Safety Function Area would have increased in color from Green to Yellow. The licensee initiated PIP M-08-6297 to address this condition. The PIP indicated that the associated work order (WO 01822572) contained comments that the switchyard work should have been deferred until the work did not have the same risk impact on Unit 1. The PIP stated that these comments were missed during the schedule review process. This failure to adequately assess risk and consequently, potentially manage risk for October 9, 2008, is a performance deficiency.

Analysis: This finding is greater than minor when compared to Inspection Manual Chapter (IMC) 0612 Appendix B, minor question 5(e) because the Unit 1 risk assessment failed to consider maintenance activities that were occurring in the switchyard that affected both units and would have resulted in a higher risk category if properly assessed. In addition, licensee risk assessment failed to consider maintenance activities that could increase the likelihood of initiating events, such as loss of offsite power. IMC 0609, Appendix M, was used to determine the safety significance because the performance deficiency was associated with a qualitative shutdown risk assessment (DID), and because Appendix K notes that it is not applicable to qualitative risk assessments. The finding was determined to be of very low safety significance (Green)

based on the IMC 0609, Appendix M evaluation. The key factor in the determination was that the reactor was defueled and time to boil in the spent fuel pool was slightly over 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. This would have allowed sufficient time such that upon a loss of offsite power there would have been a reasonable likelihood for success of actions taken to recover off-site power. In addition, after boiling started there would have been sufficient makeup water sources available to makeup to the spent fuel pool to prevent uncovering fuel.

This finding has a cross-cutting aspect of decision making in the area of human performance because the licensee did not use the information available in NSD 403 to assess the risk of the maintenance. H.1.a].

Enforcement: 10 CFR 50.65(a)(4) requires, in part, that before performing maintenance activities (including, but not limited to, surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, on October 9, 2008, the licensee failed to perform an adequate risk assessment for Unit 1, in that the overall maintenance risk assessment failed to account for activities being conducted in the switchyard which increased the likelihood of loss of off-site power. Because this violation was determined to be of very low safety significance and was placed in the corrective action program as PIP M-08-6297, this violation is being treated as a NCV in accordance with Section VI.A.1 of the Enforcement Policy: NCV 05000370/2008005-01, Failure to Perform an Adequate Risk Assessment for Switchyard Activities.

Enclosure

1R15 Operability Evaluations a. Inspection Scope For the operability evaluations listed below, the inspectors evaluated the technical adequacy of the evaluations to determine whether Technical Specification (TS)

operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors reviewed any compensatory measures taken for degraded SSCs to determine whether the measures were in-place and adequately compensated for the degradation, such that operability was justified. For the degraded SSCs, or those credited as part of compensatory measures, the inspectors reviewed the Updated Final Safety Report (UFSAR) to determine whether the measures resulted in changes to the licensing basis functions, as described in the UFSAR, and whether a license amendment was required per 10CFR 50.59. The documents reviewed during this inspection are listed in Attachment 1 of this report.

  • M-08-5649, During 1A Engineered Safeguards Feature test (concurrent safety injection/phase B//blackout section), 1A diesel generator did not meet performance standards
  • M-08-6661, 1A Control Area Ventilation and Chilled Water System (VC/YC) Chiller pulsed on during 1A Engineered Safeguards Feature test initiation, then tripped, and then restarted 15 minutes later
  • M-08-6352, oil leak on 1A charging pump motor
  • M-08-6889, Unable to obtain acceptable RN flow to 1B centrifugal charging pump oil cooler per RN flow balance Performance Test (PT)

b. Findings No findings of significance were identified.

1R19 Post-Maintenance Testing a. Inspection Scope For the eight maintenance tests listed below, the inspectors determined the safety functions described in the UFSAR and TS that were affected by the maintenance activity. The inspectors witnessed the post-maintenance tests listed and/or reviewed the test data to determine whether the test results adequately demonstrated restoration of the affected safety function(s).

  • PT/1/A/4350/002B, Diesel Generator 1B Operability Test after several planned maintenance tasks completed during the 1EOC19 refueling outage on October 16, 2008
  • PT/1/A/4206/15B, 1B Safety Injection Pump Head Curve Performance Test, conducted after 1EOC19 refueling outage maintenance.

Enclosure

  • PT/1/A/4209/12A, Centrifugal Charging Pump 1A Pump Head Curve Performance Test, conducted after 1EOC19 refueling outage maintenance
  • PT/1/A/4350/002A, Diesel Generator 1A Operability Test after several planned maintenance tasks completed during 1EOC19 refueling outage on October 23, 2008.
  • PT/1/A/4350/024, Hydrogen Mitigation Igniter Glow Plug Test, after planned replacement of glow plugs
  • PT/1/A/4252/001, Turbine Driven CA Pump Performance Test, after coupling inspection, oil change, and steam inlet valve work
  • PT/1/A/4252/001B, 1B CA Pump Performance Test after high motor vibrations and replacement of motor bearings and oil change
  • PT/0/A/4600/105, Rod Control Cluster Assembly Drop Timing Using Digital Rod Position Indicator System after re-installation and re-latching of control rods during 1EOC19 refueling outage b. Findings No findings of significance were identified.

1R20 Refueling and Outage Activities a. Inspection Scope Unit 1 began the 1EOC19 refueling outage on September 20, 2008, and remained in the outage until November 12, 2008. During this inspection period the inspectors reviewed the licensees responses to emergent work and unexpected conditions, to determine whether resulting configuration changes were controlled in accordance with the outage risk control plan. The inspectors observed outage activities and/or the items described below, to determine whether the licensee maintained defense-in-depth commensurate with the outage risk control plan for the key safety functions listed below and applicable TS when taking equipment out of service.

  • Clearance activities
  • Electrical power
  • Spent fuel pool cooling
  • Inventory control
  • Reactivity control
  • Containment closure The inspectors also observed fuel handling operations associated with core reload and other ongoing activities, to determine whether those operations and activities were being performed in accordance with TS and licensee procedures. Additionally, the inspectors observed refueling activities to determine whether the location of the fuel assemblies was tracked, including new fuel, from core offload through core reload. The inspectors observed where spent assemblies were placed to determine whether they were placed in allowable locations.

Enclosure

The inspectors assessed outage activities that were conducted during short time-to-boil periods. On October 17 and 18, 2008, the licensee chose to reduce reactor coolant level to a reduced inventory/mid-loop condition to remove steam generator nozzle dams with the new core in the reactor vessel (cold mid-loop). The inspectors reviewed the licensees commitments to GL 88-17 to determine whether they were still in place and adequate. The inspectors observed control room operations during this period to determine whether distractions were minimized.

The inspectors reviewed the licensees reactor vessel head load drop analysis and UFSAR to determine whether the UFSAR had been updated. Accordingly, the inspectors observed the setting of the reactor vessel head to determine whether procedures were followed. Prior to mode changes and on a sampling basis, the inspectors reviewed system lineups and/or control board indications to determine whether TS, license conditions, and other requirements, commitments, and administrative procedure prerequisites for mode changes were met prior to changing modes or plant configurations. Also, the inspectors periodically reviewed reactor coolant system boundary leakage data, and observed the setting of containment integrity, to determine whether the reactor coolant system and containment boundaries were in place and had integrity when necessary. Prior to reactor startup, the inspectors walked down containment to determine whether debris had been left which could affect performance of the containment sumps. The inspectors reviewed reactor startup and unit synchronization to the grid to verify procedure compliance and that the systems performed as designed. The inspectors reviewed reactor physics testing results to determine whether core operating limit parameters were consistent with the design.

When emergent rod control problems during low power physics testing resulted in the unit having to be cooled back down to Mode 5, the inspectors observed portions of the cooldown process to determine whether TS cooldown restrictions were followed.

Periodically, the inspectors reviewed the items that had been entered into the licensees corrective action program, to determine whether the licensee had identified problems related to outage activities at an appropriate threshold and had entered them into the corrective action program. The documents reviewed during this inspection are listed in Attachment 1 of this report.

b. Findings No findings of significance were identified.

1R22 Surveillance Testing a. Inspection Scope For the tests identified below, the inspectors witnessed testing and/or reviewed the test data, to determine if the SSCs involved in these tests satisfied the requirements described in the TSs, the UFSAR, and applicable licensee procedures, and that the tests Enclosure

demonstrated that the SSCs were capable of performing their intended safety functions.

The documents reviewed during this inspection are listed in Attachment 1 of this report.

Surveillance Tests

  • PT/1/A/4600/003G, Reduced Inventory Surveillance Items
  • PT/1/A/4200/009A, Engineered Safety Features Actuation Periodic Test Train A
  • PT/1/A/4200/009B, Engineered Safety Features Actuation Periodic Test Train B
  • PT/1/A/4252/016, #1 Turbine Driven CA Pump Automatic Recirculation Valve Bypass and Bleed Valve Leakage Verification In-Service Tests
  • PT/1/A/4255/03C, SM Valve (MSIV) Timing Test at Full Temperature and Pressure Containment Isolation Valve Testing
  • PT/2/A/4150/001B, Reactor Coolant Leakage Calculation Ice Condenser Systems Testing
  • PT/0/A/4200/018, Ice Bed Analysis b. Findings While reviewing Unit 1, Train A, engineering safeguard features test deficiency data on October 22, 2008, the inspectors identified that the accelerated sequencer function was not described in the UFSAR. The licensees UFSAR commits to Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Revision 1 and 3, for the format and content of the UFSAR. RG 1.70, Revision 1, Section 8.3.1.1 states to Describe the onsite A.C. power systems with emphasis placed on those portions of the systems that are safety-related. Those portions of the onsite A.C. power system that are not related to safety need only be described in sufficient detail to permit an understanding of their interactions with the safety-related portions.

The description of the safety-related portion should include: (8) automatic loading and stripping of buses. The inspectors review concluded that the accelerated sequencer function can sequentially energize various safety equipment (partitioned into load groups) from the safety-related emergency A.C. power system during design basis accidents described in UFSAR Chapter 15. This accelerated sequencer function will automatically energize the next safety load group, after 2 seconds, if the emergency A.C. bus voltage and diesel engine speed recover to values of approximately 92.5% and 97%, respectively. If after energizing certain safety load groups, the bus voltage or diesel speed permissives are no longer met, the accelerated sequencer function will drop out the start signal and those loads will become de-energized until the permissives are met again or the separate UFSAR described load sequencer function re-energizes those loads based on a timed sequence. Pending NRC review of the enforcement aspects of this issue and review of the licensees operability evaluation, this issue is identified as an Unresolved Item: URI 05000369,370/2008005-02, Accelerated Enclosure

Sequencer not Described in the UFSAR. The licensee generated PIPs M-08-6767 and M-09-0063 to address this concern.

Cornerstone: Emergency Preparedness 1EP2 Alert and Notification System Testing a. Inspection Scope The inspectors evaluated the adequacy of licensee=s methods for testing the alert and notification system in accordance with NRC Inspection Procedure 71114, Attachment 02, Alert and Notification System Evaluation. The applicable planning standard 10 CFR Part 50.47(b) (5) and its related 10 CFR Part 50, Appendix E, Section IV.D requirements were used as reference criteria. The criteria contained in NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, was also used as a reference.

The inspectors reviewed various documents which are listed in Attachment 1 of this report. This inspection activity satisfied one inspection sample for the alert and notification system on a biennial basis.

b. Findings No findings of significance were identified.

1EP3 Emergency Response Organization Augmentation a. Inspection Scope The inspectors reviewed the licensee=s Emergency Response Organization (ERO)

augmentation staffing requirements and process for notifying the ERO to ensure the readiness of key staff for responding to an event and timely facility activation. The qualification records of key position ERO personnel were reviewed to ensure all ERO qualifications were current. A sample of problems identified from augmentation drills or system tests performed since the last inspection were reviewed to assess the effectiveness of corrective actions.

The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment 03, Emergency Response Organization Staffing and Augmentation System.

The applicable planning standard, 10 CFR 50.47(b) (2) and its related 10 CFR 50, Appendix E requirements were used as reference criteria.

The inspectors reviewed various documents which are listed in Attachment 1 of this report. This inspection activity satisfied one inspection sample for the ERO staffing and augmentation system on a biennial basis.

Enclosure

b. Findings No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes a. Inspection Scope Since the last NRC inspection of this program area, Revisions 07-02 and 08-01 of the McGuire Emergency Plan were implemented based on the licensees determination, in accordance with 10 CFR 50.54(q), that the changes resulted in no decrease in the effectiveness of the Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The inspectors conducted a sampling review of the Plan changes and implementing procedure changes made between November 27, 2007, and September 17, 2008, to evaluate for potential decreases in effectiveness of the Plan. However, this review was not documented in a Safety Evaluation Report and does not constitute formal NRC approval of the changes.

Therefore, these changes remain subject to future NRC inspection in their entirety.

The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment 04, Emergency Action Level and Emergency Plan Changes. The applicable planning standard (PS), 10 CFR 50.47(b) (4) and its related 10 CFR 50, Appendix E requirements were used as reference criteria.

The inspectors reviewed various documents which are listed in Attachment 1 of this report. This inspection activity satisfied one inspection sample for the emergency action level and emergency plan changes on an annual basis.

b. Findings No findings of significance were identified.

1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies a. Inspection Scope The inspectors reviewed the corrective actions identified through the Emergency Preparedness program to determine the significance of the issues and to determine if repeat problems were occurring. The facility=s self-assessments and audits were reviewed to assess the licensee=s ability to be self-critical. In addition, the inspectors reviewed licensee=s self-assessments and audits to assess the completeness and effectiveness of all emergency preparedness related corrective actions.

The inspection was conducted in accordance with NRC Inspection Procedure 71114, Attachment 05, Correction of Emergency Preparedness Weaknesses. The applicable planning standard, 10 CFR 50.47(b) (14) and its related 10 CFR 50, Appendix E requirements were used as reference criteria.

Enclosure

The inspectors reviewed various documents which are listed in Attachment 1 of this report. This inspection activity satisfied one inspection sample for the correction of emergency preparedness weaknesses on a biennial basis.

b. Findings No findings of significance were identified.

4. OTHER ACTIVITIES 4OA1 Performance Indicator (PI) Verification

.1 Mitigating Systems Cornerstone a. Inspection Scope The inspectors sampled licensee data to confirm the accuracy of reported PI data for the indicators on Unit 1 and Unit 2 during the periods listed below. To determine the accuracy of the PI data reported during that period, the inspectors compared the licensees basis in reporting each data element to the PI definitions and guidance contained in NEI 99-02, Regulatory Assessment Indicator Guideline, Rev. 5.

  • Safety System Unavailability, Emergency AC Power
  • Safety System Unavailability, High Pressure Safety Injection
  • Safety System Unavailability, Cooling Water Systems The inspectors reviewed Licensee Event Reports, records of inoperable equipment, and Maintenance Rule records, to determine whether the licensee had adequately accounted for unavailability hours that the subject systems had experienced during the previous four quarters (October 1, 2007 - September 30, 2008). The inspectors also reviewed the number of hours those systems were required to be available and the licensees basis for identifying unavailability hours. In addition, the inspectors interviewed licensee personnel associated with the PI data collection, evaluation, and distribution.
  • Safety System Functional Failures The inspectors reviewed Licensee Event Reports and Maintenance Rule records, to determine whether the licensee had adequately accounted for functional failures that the subject systems had experienced during the previous four quarters (October 1, 2007 -

September 30, 2008).

b. Findings No findings of significance were identified.

Enclosure

.2 Emergency Preparedness Cornerstone a. Inspection Scope The inspectors sampled licensee submittals for the three PIs listed below. For each of the submittals reviewed, the inspectors reviewed the period from July 1, 2007, through June 30, 2008. To verify the accuracy of the PI data reported during that period, PI definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Indicator Guideline, Revision 5, were used to verify the basis in reporting for each data element.

  • Emergency Response Organization Drill/Exercise Performance (DEP)
  • Emergency Response Organization Readiness (ERO)
  • Alert and Notification System Reliability (ANS)

The inspectors reviewed portions of the raw PI data developed from monthly performance indicator reports and discussed the methods for compiling and reporting the PIs with cognizant emergency preparedness personnel. The inspectors also independently screened drill and exercise opportunity evaluations, drill participation reports, and drill evaluations. Selected reported values were calculated to verify their accuracy. The inspectors compared graphical representations from the most recent PI report to the raw data to verify that the data was correctly reflected in the report.

Reviewed documents are listed in Attachment 1 of this report.

b. Findings No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Routine Review 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 screening of items entered into the licensees corrective action program. This was accomplished by reviewing copies of condition reports, attending some daily screening meetings, and accessing the licensees computerized database.

.2 Selected Issue Follow-Up Inspection-Operator Workaround a. Inspection Scope The inspectors selected PIP M-08-4774 associated with the 2A emergency diesel generator failing to start during surveillance testing for detailed review. The inspectors reviewed this report to determine whether the licensee identified the full extent of the issue, performed an appropriate evaluation, and specified and prioritized appropriate corrective actions. The inspectors evaluated the licensee documents against the Enclosure

requirements of the licensees corrective action program and 10 CFR 50, Appendix B.

The documents reviewed during this inspection are listed in Attachment 1 of this report.

The inspectors also performed a review of the 14 priority 1-3 operator workarounds (OWAs) listed in the licensees December 2008 OWA report to determine whether the OWAs were identified in the corrective action program and whether corrective actions have been properly identified and dates established for completion. In some cases the review included the PIPs associated with the OWA and a review of the system health report for the associated system. A review of selected workarounds closed in the last 2 years was conducted to determine whether the closed OWAs were corrected.

b. Findings No findings of significance were identified.

.3 Semi-Annual Review to Identify Trends a. Inspection Scope The inspectors performed a trend review to determine if trends were identified outside the corrective action program that could indicate the existence of a more significant safety issue. The inspectors review was focused on repetitive equipment issues, but also considered the results of daily inspector corrective action program item screening discussed above, licensee trending efforts, and licensee human performance results.

The review also included issues documented outside the normal Corrective Action Program in major equipment problem lists, plant health team vulnerability lists, focus area reports, system health reports, self-assessment reports, maintenance rule reports, and Safety Review Group Monthly Reports. The inspectors compared and contrasted their results with the results contained in the licensees latest quarterly trend reports.

Corrective actions associated with a sample of the issues identified in the licensees trend report were reviewed for adequacy. The documents reviewed during this inspection are listed in Attachment 1 of this report.

b. Findings and Observations Update of previously identified trends The inspectors previously identified a trend associated with numerous violations for failing to update the FSAR in accordance with regulations outlined in 10 CFR Part 50.71(e). The licensee initiated PIP M-06-2889 to address the UFSAR accuracy trend and performed a sample review of the UFSAR. In addition, the licensee identified a number of UFSAR problems from their UFSAR review program during this inspection period. During the 6 month period, an additional example of UFSAR inaccuracies was identified by the NRC. NRC identified that the UFSAR had not been updated to reflect the fact that an additional load sequencer function, the accelerated sequencer function, was not described in the UFSAR as discussed in section 1R22 of this report. This example was outside the scope of the licensees UFSAR review process, because it was Enclosure

due to an inaccuracy of the FSAR at the time of licensing. PIP M-08-4383 (RN sharing violation) also evaluated the corrective action for previous UFSAR PIPs associated with violations and whether the corrective actions were effective and adequate. Additional programmatic changes are being recommended from the apparent cause identified in PIP M-08-4383.

New trends No new trends were identified this period that had not already been identified by the licensee.

4OA3 Event Follow-up 1. Rod Control Urgent Failure Alarm a. Inspection Scope On October 31, 2008, the inspectors reviewed the licensees actions associated with a rod control urgent failure alarm while testing control bank B during reactor startup physics testing (procedure PT/0/A//4150/028, Initial Criticality and Zero Power Physics Testing). The reactor was at 10-8 amps on the intermediate range nuclear instrumentation. The licensee entered AP/1/A/5500/ 012, Rod Control Malfunction, and during actions in the Abnormal Procedure (AP), control rod K-2 dropped (control bank B, Group II). The licensee shutdown the reactor as required by the AP in accordance with procedure OP/1/A/6100/003, Controlling Procedure for Unit Operation, which required a shutdown by manual reactor trip. The inspectors observed and evaluated performance of the plant systems and the operators including the proper use and adherence to plant procedures. The inspectors also reviewed the licensees classification and reporting of the event. The event was reported on December 30, 2008, as Licensee Event Report 05000369/2008-03.

On November 3, 2008, as part of post-maintenance testing for the previous control rod concern conducted with the plant in a cold shutdown condition, another control rod urgent failure alarm was received and operators opened the reactor trip breakers to insert all remaining control rods. The inspectors reviewed the licensees initial notification of the event and subsequent retraction of that notification. Documents reviewed during this inspection are listed in Attachment 1 of this report.

b. Findings No findings of significance were identified.

Enclosure

.2 Unit 1 startup after refueling a. Inspection Scope The inspectors also reviewed the licensees actions associated with the Unit 1 approach to criticality on October 31, 2008, and November 12, 2008, as well as power escalation on November 13, 2008. The inspectors evaluated the performance of the operators and assessed procedural compliance.

b. Findings No findings of significance were identified.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities a. Inspection Scope During the inspection period, the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours. These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status review and inspection activities.

b. Findings No findings of significance were identified.

.2 (Closed) URI 05000370/2007005-02, Failure to Control a Locked-High Radiation Area (LHRA) Barrier a. Inspection Scope Following an investigation by NRCs Office of Investigations (OI) to determine willfulness, the inspectors performed an in-office review to appropriately disposition URI 05000370/ 2007005-02. Documents reviewed are listed in Attachment 1 of this report.

b. Findings Introduction: A Severity Level (SL) IV NCV of TS 5.7.2 was identified for failure to control access to a LHRA when the Unit 2 reactor head inspection stand LHRA barrier was left unlocked and unguarded.

Enclosure

Description: On September 30, 2006, two AREVA technicians were tasked with moving the reactor head inspection manipulator out from under the reactor head (the LHRA) to outside of the reactor head inspection stand shield and replacing the dummy sleeve on the manipulator. The work area was located inside a highly contaminated area (HCA)

which encompassed the reactor head inspection stand.

During a pre-job brief for the work, a contract RPT was assigned to have the LHRA key and procedure SH/0/B/2000/012, Enclosure 5.8 (Accessing Extra High Radiation Area <

10 R/hr) and work outside of the HCA.

Step 10 in procedure SH/0/B/2000/012, Enclosure 5.8, requires double verification that the padlock is secured by pushing/pulling on the padlock and double verification that the access (e.g., door, gate, etc.) is secure. A second radiation protection technician was assigned to work inside the HCA and monitor the AREVA technicians activities.

Upon beginning work, the AREVA technicians drove the manipulator out from under the reactor head inspection stand into the HCA to setup for the work. The technician providing job coverage inside the HCA closed the upper and lower doors to the reactor head inspection stand barrier, but could not completely close or lock the doors because the manipulator physically prevented the lower doors from completely closing. A lead blanket was placed against the upper doors to provide additional shielding covering the upper and lower doors latching and locking mechanisms.

Prior to completion of the work, both AREVA technicians and the job coverage technician were required to leave containment. The work was left with the reactor head inspection stand LHRA barrier upper doors pushed closed, but neither latched nor locked. The status/configuration of the barrier (including the lock) could not be verified from where the RPT responsible for controlling the barrier was located outside of the HCA. The doors to the reactor head inspection stand appeared to be closed, although in a configuration contrary to what was discussed during the pre-job brief, the scaffold pin was through the handles, and a lead blanket was over the area where the lock would have been.

The RPT assumed that the job coverage technician had locked the LHRA. The RPT then notified the RP Lead Technician that he needed to use the restroom. While in the restroom, the RPT realized the LHRA door may not have been locked. Furthermore, he realized that he had not tested the lock (pushed and pulled the locking mechanism) as written in the procedure. As a consequence, from approximately 5:05 am to 5:21 am, the Unit 2 reactor head inspection stand LHRA was left unlocked and unguarded subsequent to workers leaving the area.

Analysis: The significance of the violation was assessed using traditional enforcement, because it involved willfulness (see OI Synopsis in Attachment 2 of this report). In accordance with Supplement IV, Health Physics, of the NRC Enforcement Policy, the NRC determined that the safety significance of this violation was SL IV because the situation described in example 7 of a SL III violation (the finding involves a situation with a substantial potential for exposure in excess of applicable limits) did not exist and, per example 9 of a SL IV violation, was a matter with more than a minor safety, health, or Enclosure

environmental significance. Although this violation involved willfulness, it was dispositioned as an NCV in accordance with Section IV.A.1 of the Enforcement Policy because the licensee identified the violation and promptly discussed it with regional health physics inspectors, the violation involved the acts of a low-level individual, the violation appears to be the isolated action of the employee without management involvement, and significant remedial action commensurate with the circumstances was taken by the licensee.

Enforcement: TS 5.7.2 requires, in part, that areas with radiation levels greater than 1000 mrem/hr at 30 cm (12 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked or continuously guarded doors to prevent unauthorized entry. Contrary to the above, the Unit 2 reactor head inspection stand locked high radiation area barrier was left unlocked and unguarded for approximately 16 minutes on September 30, 2006, when the RPT responsible for guarding/locking the barrier left containment to use the restroom.

The finding was documented in the licensees corrective action program as PIP M-06-4479. Because this violation was of very low safety significance and was entered into the licensees corrective action program, it is being treated as an NCV, consistent with Section IV.A of the NRC Enforcement Policy: NCV 05000370/2008005-03, Failure to Control a Locked-High Radiation Area Barrier [EA-08-268]. As such, URI 05000370/2007005-02 is considered closed.

.3 (Closed) NRC Temporary Instruction (TI) 2515/175, Emergency Response Organization, Drill/Exercise Performance Indicator, Program Review The inspectors completed Temporary Instruction TI 2515/175, Emergency Response Organization, Drill/Exercise Performance Indicator, Program Review. Appropriate documentation of the results was provided to NRC, HQ, as required by the TI. This completes the Region II inspection requirements for this TI for McGuire Nuclear Station.

4OA6 Meetings, Including Exit On January 13, 2009, the resident inspectors presented the inspection results to Mr. B.

Hamilton, Oconee Site Vice President, and other members of his staff. The inspectors confirmed that proprietary information was not provided or examined during the inspection. No proprietary information is included in this report.

4OA7 Licensee-Identified Violations None ATTACHMENTS: (1) SUPPLEMENTAL INFORMATION (2) OI SYNOPSIS Enclosure

SUPPLEMENTAL INFORMATION KEY POINTS OF CONTACT Licensee Ashe, K., Manager, Regulatory Compliance Black, D., Security Manager Bradshaw, S., Training Manager Branch, R., Inservice Inspection Manager Brewer, D., Manager, Safety Assurance Capps, S., Manager, Engineering Crane, K., Regulatory Compliance Cutri, G., Primary Engineer Downing, P., Manager, Steam Generator Hamilton, B. Site Vice President, McGuire Nuclear Station Hatley, M., Steam Generator Engineer Hicks, J., Superintendent, Maintenance Murray, K., Emergency Preparedness Manager Mooneyhan, S., Radiation Protection Manager Nolin, J., Manager, Mechanical and Civil Engineering Parker, R., Superintendent, Work Control Repko, R., Station Manager, McGuire Nuclear Station Scott, W., Chemistry Manager Sheffield, R, ISI/NDE Simril, T., Superintendent, Plant Operations Snider, S., Manager, Reactor and Electrical Systems Engineering Swann, C., BACCP Engineer Underwood, G, NGO NRC personnel J. Stang, Project Manager, NRR R. Carroll, Senior Project Engineer, RII S. Rose, Acting Branch Chief, RII LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Open 05000369,370/2008005-02 URI Accelerated Sequencer not Described in the UFSAR (Section 1R22)

Attachment 1

Opened and Closed 05000370/2008005-01 NCV Failure to Perform an Adequate Risk Assessment for Switchyard Activities (Section 1R13)05000370/2008005-03 NCV Failure to Control a Locked-High Radiation Area Barrier [EA-08-268] (Section 4OA5.2)

Closed 05000370/2007005-02 URI Failure to Control a Locked-High Radiation Area Barrier (Section 4OA5.2)

2515/175 TI Emergency Response Organization, Drill/Exercise Performance Indicator, Program Review (Section 4OA5.3)

DOCUMENTS REVIEWED Section 1R01: Adverse Weather Protection IP/1/B/3250/059B, Monthly Check of Freeze Protection IP/2/B/3250/059B, Monthly Check of Freeze Protection IP/0/B/3250/059C, Preventative Maintenance and Operational Check of Freeze Protection for Intake PT/0/B/4700/038, Verification of Freeze Protection Equipment and Systems PT/0/B/4700/070, On Demand Freeze Protection Verification Checklist Section 1R04: Equipment Alignment Partial System Walkdown

[Emergency Diesel Generator 1A]

Drawing MCFD-1609-04.00, Flow Diagram of the Diesel Generator Starting Air System Drawing MCFD-1609-03.00, Flow Diagram of the Diesel Generator Engine 1A Fuel Oil System Drawing MCFD-1609-02.00, Flow Diagram of the Diesel Generator Engine Lube Oil System Drawing MCFD-1609-01.00, Flow Diagram of the Diesel Generator Engine Cooling Water System

[Emergency Diesel Generator 2A]

Drawing MCFD-2609-04.00, Flow Diagram of the Diesel Generator Starting Air System Drawing MCFD-2609-03.00, Flow Diagram of the Diesel Generator Engine 2A Fuel Oil System Drawing MCFD-2609-02.00, Flow Diagram of the Diesel Generator Engine Lube Oil System Drawing MCFD-2609-01.00, Flow Diagram of the Diesel Generator Engine Cooling Water System Attachment 1

[Safety Injection System 1A]

Drawing MCFD-1562-03.00, Flow Diagram of Safety Injection System (NI)

Drawing MCFD-1562-03.01, Flow Diagram of Safety Injection System (NI)

Complete Walkdown (SSF):

OP/0/B/6350/004, Standby Shutdown Facility Diesel Operation PT/0/A/4200/002, Standby Shutdown Facility Operability Test UFSAR sections: 8.4 & 9.5.1 NUREG-0422 and supplements, Safety Evaluation Report related to operation of McGuire Nuclear Station, Units 1 and 2 MCS-1223.SS-00-0001, Design Basis Specification for the Standby Shutdown System SSF System Health Report Auxiliary Feedwater System Health Report Nuclear Service Water System Health Report PIPs: M-06-2284, M-07-6044, M-07-6079, M-08-7600 PIPs initiated from this inspection: M-09-00103, Evaluate UFSAR content for 10CFR50.63 with respect to valve 0RN-4AC Section 1R05: Fire Protection McGuire Nuclear Station IPEEE Submittal Report dated June 1, 1994 McGuire Nuclear Station Supplemental IPEEE Fire Analysis Report dated August 1, 1996 MCS-1465.00-00-0008, Design Basis Specification for Fire Protection Section 1R08 Inservice Inspection (ISI) Activities Procedures and Specifications SGMEP 105, McGuire Unit 1 CFR-80 Specific Assessment of Potential Degradation Mechanisms for 1EOC19, Rev. 7 Brook Associate Procedure 83-0095, Remote Examination & Removal of Foreign Objects from Steam Generator Secondary Side, Rev. 0 Areva Procedure 03-1246524, Instruction for Plug Inspection, Rev. 009 Areva Procedure 03-1275114, Eddy Current Data Management Guidelines, Rev. 012 Areva Procedure 03-1277368-012, System Administration Guidelines for Eddy Current Assignments Areva Procedure 03-6037041, Standalone Steam Generator Machine Vision System Field Instructions, Rev. 004 Areva Procedure 54-ISI-400-16, Multi-Frequency Eddy Current Examination of Tubing, April 1, 2008 Areva Procedure 03-9072867, Eddy Current Guidelines for Duke Energy Companys CFR-80 Steam Generators, Rev. 003 MP/1/A/7150/042 A, Rx Vessel Head Removal, Rev. 16 (Sections 11.4.2-11.4.7, 11.13.1-11.13.7 MP/0/A/7700/080, Inspection and Cleanup of Boric Acid on Plant Materials, Rev. 11 MP/0/A/7650/034, Fabrication and Erection of Structural and Miscellaneous Steel, Rev. 23 PT/0/A/4150/046, Containment Walkdown, Rev. 3 NDE 12, General Radiography Procedure for Preservice and Inservice Inspection, Rev. 12 Attachment 1

NDE 35, Liquid Penetrant Examination, Rev. 21 NDE 680, Ultrasonic Examination of Nozzle Inner Radii in Ferritic Pressure Vessels, Rev. 5 NGD Welding Manual, Functional Area Manual, Guidelines for Controlling Station Welding and Associated Processes, Rev. 12 NSD 300, ASME Section XI Program, Rev. 7 NSD 322, Boric Acid Corrosion Program, Rev. 1 NSD 400, Nuclear Generation Welding Program, Rev. 5 NSD 603, Special Processes, Rev. 5 NSD 702, Document Control, Rev. 20 NSD 703, Administrative Instruction for Technical Procedures, Rev. 20 Duke Energy Carolinas Topical Report, Quality Assurance Program, Amendment 36 ASME Section IX Welding Program - Program Manual, Rev. 10 Welding Procedure Specification GTSM0808-01, Rev. 0 - Rev. 6 Procedure Qualification Records: L-110D, L-112, L-108, L-148C, L-138, L-109 Weld Process Control for weld 1NS88A-10 PIPs C-03-5744, C-06-0874, C-08-3199, G-08-0152, M-01-3731, M-02-3415, M-06-3438, M-06-3772, M-06-4832, M-06-5169, M-06-5548, M-07-158, M-07-2002, M-07-2747, M-07-4405, M-07-5327, M-08-3147, M-08-5901, M-08-5924, M-08-5928, M-08-5938, M-08-5990 Other Records Duke, CFR-80 Steam Generator Site Technique Validation for Catawba Unit 1 and McGuire Units 1 and 2, Rev. 7 Eddy Current Examination Technique Specification Sheets Work Order 589722, Foreign Objects Inspection and Retrieval for SG A Personnel Qualification and Certification Records Equipment and Calibration Certification Records Tube Data Acquisition and Analysis Records Process Control Sheets for Foreign Object Log Work Order (WO) 01714760 01, Unit 1 Refueling Water Storage Tank WO 01714760 03, Unit 1 Refueling Water Storage Tank WO 01698701 01, Hanger 1-MCR-NC-0579 WO 01698687, UT examination CAL-07-014 CAL-07-015 Inservice Inspection Report, McGuire Unit 1, 2007 Refueling Outage, EOC 18 Third Interval Inservice Inspection Plan, McGuire Nuclear Station Units 1 and 2, General Requirements, Rev. 2 Boric Acid Corrosion Control Program Assessment, November 13-15, 2007 Meeting Minutes from Welding B.E.S.T., February 27, 2008 Minor Modification, ME100776, Cut out and Replace 8, Schedule 10 piping with 8 Schedule 40 Section 1R11: Licensed Operator Requalification AP/1/A/5500/06, Loss of Feedwater AP/1/A/5500/05, Generator Voltage and Electrical Grid Disturbances Attachment 1

AP/1/A/5500/03, Load Rejection AP/1/A/5500/10, NC System Leakage within the Capacity of Both NV Pumps AP/1/A/5500/014, Rod Control Malfunction EP/1/A/5500/E-0, Reactor Trip or Safety Injection EP/1/A/5500/E-3, Steam Generator Tube Rupture EP/1/A/5500/F-0, Critical Safety Function Status Trees EP/1/A/5500/ES-0.1, Reactor Trip Response, Section 1R12: Maintenance Effectiveness M-08-4781 M-08-5757 M-07-00877 M-07-04953 McGuire Work Order #1833494 PT/2/A/4350/002A, Diesel Generator 2A Operability Test-Section1R13: Maintenance Risk Assessments and Emergent Work Evaluation Nuclear System Directive 403, Shutdown Risk Management (Modes 4, 5, 6, and No-Mode) per 10 CFR 50.65(a)(4)

MD201629, upgrade of 2B RN strainer dP instrumentation Section1R15: Operability Evaluations IP/1/A/3250/012A, Diesel Load Sequencer 1A Timer Calibration, performed on 3/29/07 UFSAR Section 8 PT/1/A/4350/019A, 1A D/G /Governor and Voltage Regulator Benchmark Comparison Test, performed on 10/4/2008 TS 3.8.1 and 3.8.2 PT/1/A/4200/009A, Engineered Safety Features Actuation Periodic Test Train A PT/1/A/4350/002A 1A D/G Operability Test PIP M-08-5672 NRC RIS 2005-20 NSD 203, Operability, Functionality TS 3.7.9 and 3.7.10, and their bases UFSAR sections 8, 7.6.10, and 9.4 PT/1/A/4200/009A, Engineered Safety Features Actuation Periodic Test Train A MCS-1578.VC-00-001, Design Basis Specification for the VC/YC System PIP M-08-5632 RG 1.70 10CFR50.59 evaluation for MD500739and MD500740 PIPs generated from this IP: M-08-6767 ASME Code Case N-513-2 PIP M-08-6775 RIS 2005-20. R1 NSD 203, Operability, Functionality Attachment 1

RG 1.147 UFSAR section 9.2.2, and Table 9-8 TS 3.7.7 and bases SLC 16.5-9 PM 1784692 PIPs M-07-4758 and M-08-1080 Section1R20: Refueling and Outage Activities MP/1/A/7150/042B, Reactor Vessel Head Installation IP/0/B/3262/001, Overhead Cranes and Hoists Electrical Inspection and Maintenance MP/0/A/7700/096, Quarterly/Annual Inspection and Servicing of Overhead and Gantry Cranes MP/0/A/7150/136, Inspection of Reactor Vessel Head and Internals Lift Rigs MP/0/A/7150/110, Control Rod Drive Mechanism Missile Shield Lifting Rig Inspection MP/1/A/7650/060, Operation of Polar crane in Unit 1 Upper Containment Calculation MCC-1134.02-00-0043, Reactor Vessel Head Drop Analysis UFSAR section 9.1.5 PIPs related to Rx head lift: M-07-3099, M-07-4759, and M-07-5268 OP/1/A/6100/SU-2, Refueling and Replacing Reactor Vessel Head OP/1/A/6100/SO-1, Maintaining NC System Level OP/1/A/6100/SO-3, Draining the Refueling Cavity Generic Letter 88-17, Loss of Decay Heat Removal Licensee responses to GL 88-17 dated 1/3/1989, 2/2/1989,3/10/1989, 9/12/1991, and 10/16/1991, SLCs 16.5.1, 16.5.2,, 16.5.3, 16.5.4, 16.5.5; Associated with Reduced Inventory Operation OP/1/A/6100/SU-3, Mode 5 Checklist MP/0/A/7150/076 Ice Basket Weight Determination OP/1/A/6100/003, Controlling Procedure for Unit Operation PT/0/A/4150/021, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, &

Power Escalation Testing PT/0/A/4150/028, Initial Criticality and Zero Power Physics Testing OP/1/A/6100/SD-1, Prepare for Cooldown OP/1/A/6100/SD-2, Cooldown to 400 Degrees F OP/1/A/6100/SD-4, Cooldown to 240 Degrees F OP/1/A/6100/SD-6A, Placing Train A Residual Heat Removal in Service OP/1/A/6100/SD-6B, Placing Train B Residual Heat Removal in Service OP/1/A/6100/SD-7, Cooldown to 200 Degrees F OP/1/A/6100/SO-10, Controlling Procedure for LTOP Operation Section1R22: Surveillance Testing 10 CFR Part 50, Appendix J PT/1/A/4200/044, Containment Structural Integrity Inspection Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J ANSI/ANS-56.8-1994, Containment System Leakage Testing Requirements Attachment 1

Section 1EP2: Alert and Notification System Testing Procedures and Documentation McGuire Nuclear Station Site-Specific Offsite Radiological Emergency Preparedness Alert and Notification System Quality Assurance Verification, September 12, 1986 Report to FEMA - McGuire Alert & Notification System, September 10, 1984 PT/0/A/4600/103C, Siren System Annual Preventive Maintenance Review, Rev. 2 EPFAM (Emergency Planning Functional Area Manual) 3.3, Alert and Notification System (Siren Program), Rev. 9 255232M1, 2001 Siren Federal Signal Corporation - Installation and operating instructions, Rev.

M1 Records and Data Siren system availability test records for October 2006 to June 2008 Annual Preventive maintenance documentation for 2006, 2007, 2008 PIPs M-08-6260, Siren ground values M-08-6250, Siren test October 8, 2008 Section 1EP3: Emergency Response Organization Augmentation Procedures PT/0/A/4600/97, Procedure for Preparing and Conducting Emergency Exercises/Drills, Rev. 8 MTP 7111.0, Emergency Response (ER) Training Program, Rev. 8 Records and Data McGuire Emergency Response Organization Chart, October 7, 2008 Drill package for 2007 Augmentation Drill, August 15, 2007 at 1855 hrs Drill package for 2008 Augmentation Drill, August 6, 2008 at 1901 hrs Summary Data on failures in weekly functional test of ERO pagers, September 2007 -

September 2008 Training records for multiple personnel were reviewed Section 1EP4: Emergency Action Level and Emergency Plan Changes Plans and Changes packages Emergency Plan 07-02, December 12, 2007 Emergency Plan 08-01, September 25, 2008 HP/0/B/1009/029, Initial Response On-Shift Dose Assessment, Rev. 009 Notification to the State and Counties from the, Rev. 15, and Rev. 16 RP/0/A/5700/011, Conducting a Site Assembly, Site Evacuation or Containment Evacuation, Rev. 011 RP/0/A/5700/018, Notification to the State and Counties from the Technical Support Center, Rev. 019 SR/0/B/2000/004, Notification to the State and Counties from the Emergency Response Facility for Catawba, McGuire, and Oconee, Rev. 13 Attachment 1

Section 1EP5: Correction of Emergency Preparedness Weaknesses and Deficiencies Procedures NSD 208, Problem Investigation Process, Rev. 30 Audits and Self-Assessments Self-Assessment No. EP-SA06-04, Emergency Planning Corrective Action Effectiveness Review Self-Assessment No. EP-SA07-01, Siren Remote Controller (RC) Preventive Maintenance (PM)

Effectiveness Self-Assessment No. EMP-SA08-01, Assessment Title: 50.54(q) Quality and Spartanburg Training Effectiveness Self-Assessment No. EMP-SA08-02, Emergency Notification Forms Self-Assessment No. EP-SA07-03, Current EAL Scheme Compliance with RIS 2007-01(Clarification of NRC Guidance for Maintaining A Standard Emergency Action Level Scheme GO-06-019(NPA) (EP) (ALL), 2006 Emergency Planning Functional Area Assessment, December 5, 2006 GO-07-20(NPA) (EP) (ALL), 2007 Emergency Planning Functional Area Assessment, November 12, 2007 08-08(INOS) (EP) (MNS), 2008 Emergency Planning Regulatory Program Audit Records and Data Emergency Medical Drill, November 9, 2006 at 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br /> 2007-03-D2, Training Drill for Team D, August 29, 2007 2007-06-C1, Training Drill for Team C, December 5, 2007 2008-02-E3, Training Drill for Team E, May 7, 2008 2008-03-B4, Training Drill for Team B, August 27, 2008 PIPs M-06-05436, 2006 Annual Contaminated Injury Drill M-06-05679, ERO Training Drill 2006-05-E3 M-06-05830, TSC elevated temperatures M-06-05830, TSC Ventilation Work Order Priority M-07-00012, ANS lost communication M-07-01509, ANS Radio frequency Interference (RFI) noise M-07-02603, ANS Lincoln County Siren Activation M-07-03471, ERO Training Drill 2007-02-A5 M-07-04330, ANS Interference M-08-00083, Job Performance Measures used for Performance Indicator Data Section 4OA1.2: PI Verification - EP Procedures Emergency Planning Functional Area Manual Section 3.7 - NRC Regulatory Assessment Performance Indicator Guideline - Emergency Preparedness Cornerstone, Rev. 14 Attachment 1

Records and Data ANS data from 3rd Qtr 2007 to 2nd Qtr 2008 DEP data from 3rd Qtr 2007 to 2nd Qtr 2008 ERO data from 3rd Qtr 2007 to 2nd Qtr 2008 Section 4OA2: Identification and Resolution of Problems McGuire Work Order #1779566 MCID-2499-ZD.01 instrument detail for diesel crankcase vacuum pressure switch PT/2/A/4350/002A, Diesel Generator 2A Operability Test PIP M-08-7400 PIP M-08-7385 PIP M-08-2639 PIP M-08-2765 PIP M-08-4153 PIP M-08-5507 PIP M-08-7180 UFSAR violation history 2004003-02 2 examples, SSF and FW isolation valve stroke timing 2004005-02 RN Train A swap to pond procedure not described (50.59)

2004005-03 Changed # of required trains of AFW in UFSAR to less than TS (50.59)

2005004-01 Inadequate corrective action (XVI) for not updating UFSAR for SSF 2005004-02 UFSAR not updated for CAPRM amendment 2006004-02 Failure to adequately update UFSAR for SSF 2006004-03 UFSAR not updated for station blackout 2006005-02 XVI for U2 sump debris/Throttle Valves (2007008 AV, 010 NCV)

2007003-01 XVI for U1 sump debris/Throttle Valves (2007005-04 NCV)

2007003-02 Reactor vessel head lift analysis not in UFSAR (enforcement discretion)

2007004-01 Removing Ice fusion from UFSAR (50.59)

2008003-02 UFSAR not updated for RN sharing between units amendment Section 4OA3: Event Follow-up E-0, Reactor Trip or Safety Injection ES-0.1, Reactor Trip Response AP/1/A/5500/14 Rod Control Malfunction PT/1/A/4600/001 RCCA Movement Test Section 4OA5.2: URI 05000370/2007005-02 Procedures, Instructions, Guidance Documents, and Operating Manuals SH/0/B/2000/003, Preparation of a Radiation Work Permit, Rev. 7 SH/0/B/2000/012, Access Controls for High, Extra High, and Very High Radiation Areas, Rev. 7 Attachment 1

Records and Data Radiation Work Permit 2060, U2 Reactor Building, UT and Volumetric Testing Under the Reactor Head, Rev. 2 Survey # M-100306-30, U2 Rx Head Inspection Staging Area Survey # M-093006-28, U2 Rx Head Inspection Staging Area Corrective Action Program Documents PIP M-06-4479, Loss of control of EHRA barrier, 10/2/06 Root Cause Failure Analysis Report, PIP M-06-4479 LIST OF ACRONYMS ANS - Alert and Notification System AP - Abnormal Procedure ASME - American Society of Mechanical Engineers BACC - Boric Acid Corrosion Control CA - Auxiliary Feedwater CVCS - Chemical Volume Control System DEP - Drill/Exercise Performance DID - Defense-in-Depth EDG - Emergency Diesel Generator EOC - End-of-Cycle ERO - Emergency Response Organization IMC - Inspection Manual Chapter ISI - In-Service Inspection HCA - Highly Contaminated Area LHRA - Locked High Radiation Area NCV - Non-Cited Violation NDE - Non-Destructive Examination NEI - Nuclear Energy Institute OI - Office of Investigations OP - Operating Procedure OWA - Operator Workaround PI - Performance Indicator PIP - Problem Investigation Process report PS - Planning Standard PT - Performance Test RN - Nuclear Service Water RPT - Radiation Protection Technician SG - Steam Generator SSC - Structures, Systems and Components SSF - Standby Shutdown Facility TI - Temporary Instruction TS - Technical Specifications UFSAR - Updated Final Safety Analysis Report UT - Ultrasonic Test VC/YC - Control Area Ventilation and Chilled Water System VT - Visual Examination WO - Work Order Attachment 1

OI SYNOPSIS This investigation was initiated by the U.S. Nuclear Regulatory Commission, Office of Investigations, Region II, on December 19, 2007, to determine whether a Radiation Protection Technician employed by Bartlett Nuclear Services at the McGuire Nuclear Station willfully failed to ensure that a Locked High Radiation Area was properly locked and controlled prior to leaving containment.

A separate element of the concern involved the inability of Duke Energy Corporation to locate and provide the pre-job brief form and procedure steps with double verification sign-off of the LHRA barrier to the NRC. Information related to this issue is addressed in the supplemental section of the OI report.

Based on the evidence developed during this investigation, OI: Region II substantiated that the RPT for Bartlett willfully failed to follow the procedural requirements for a Locked High Radiation Area contained in the McGuire Technical Specification 5.7.2 (failure to control a LHRA barrier)

and ensure that a LHRA was properly locked and controlled prior to leaving containment.

Attachment 2

Retrieved from "https://kanterella.com/w/index.php?title=IR_05000369/2008005&oldid=2149570"