IR 05000397/2009003

IR 05000397-09-003 on 03/29/2009 - 06/27/2009; Columbia Generating Station; Integrated Resident and Regional Report; Access Control to Radiologically Significant Areas; Identification and Resolution of Problems

ML092220243
Person / Time
Site:	Columbia
Issue date:	08/10/2009
From:	Webb Patricia Walker NRC/RGN-IV/DRP/RPB-A
To:	Parrish J Energy Northwest
References
IR-09-003
Download: ML092220243 (46)

Text

August 10, 2009

Mr. J. Chief Executive Officer Energy Northwest P.O. Box 968, Mail Drop 1023 Richland, WA 99352-0968

Subject:

COLUMBIA GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000397/2009003

Dear Mr. Parrish:

On June 27, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Columbia Generating Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on July 7, 2009, with Mr. S. Oxenford, Vice President, Nuclear Generation and other members of your staff.

The 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 two self-revealing findings of very low safety significance (Green). One of these findings was determined to involve a violation of NRC requirements. However, because of the very low safety significance and because this is entered into your corrective action program, the NRC is treating this finding as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest the violation or the significance of the noncited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Columbia Generating Station facility. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector at the Columbia Generating Station. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.

UNITED STATES NUCLEAR REGULATORY COMMISSION R E GI ON I V 612 EAST LAMAR BLVD, SUITE 400 ARLINGTON, TEXAS 76011-4125

Energy Northwest

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 Room or from the Publicly Available Records component of NRCs document system (ADAMS).

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

Sincerely,

/RA/

Wayne C. Walker, Chief Project Branch A Division of Reactor Projects

Docket: 50-397 License: NPF-21

Enclosure:

NRC Inspection Report 05000397/2009003 w/Attachment: Supplemental Information

REGION IV==

Docket:

50-397 License:

NPF-21 Report:

05000397/2009003 Licensee:

Energy Northwest Facility:

Columbia Generating Station Location:

Richland, Washington Dates:

March 29, 2009 through June 27, 2009 Inspectors:

R. Cohen, Senior Resident Inspector M. Hayes, Resident Inspector N. Hernandez, Project Engineer E. Ruesch, Reactor Inspector C. Graves, Health Physicist B. Henderson, Reactor Inspector J. Adams, Reactor Inspector Approved By:

W. Walker, Chief, Project Branch A Division of Reactor Projects

- 2 -

Enclosure

SUMMARY OF FINDINGS

IR 05000397/2009003; 03/29/2009-06/27/2009; Columbia Generating Station; Integrated

Resident and Regional Report; Access Control to Radiologically Significant Areas; Identification and Resolution of Problems

The report covered a 3-month period of inspection by resident inspectors and announced baseline inspections by regional based inspectors. One Green noncited violation and one Green finding of significance were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, Significance Determination Process. Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

Cornerstone: Initiating Events

Green.

The inspectors reviewed a self-revealing finding for the failure of Energy Northwest to implement the standards and guidance provided in Site Wide Procedure SWP-CAP-01, Corrective Action Program, Revision 17. Specifically,

Energy Northwest failed to take prompt corrective action in response to Action Request 1485, dated September 2000, that identified the Cuno filter as a single point vulnerability, which could lead to a plant scram. Action Request 1485 recommended upgrading the type of filter in the seal oil system to a high efficiency duplex filter assembly. Due to a low priority ranking, corrective action was delayed several times. Action Request 1485-4, dated March 11, 2008, documented a scheduling error delaying the corrective action from fiscal year 2009 to fiscal year 2010 or 2011.

The finding was more than minor because it affected the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609.4, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The finding had a crosscutting aspect in the area of problem identification and resolution associated with operating experience because the licensee failed to implement operational experience through changes to station processes, procedures, equipment, and training programs

[P.2.(b)](Section 4OA2).

Cornerstone: Occupational Radiation Safety

Green.

A self-revealing noncited violation of Technical Specification 5.7.1 was identified for failure to barricade and conspicuously post a high radiation area.

On April 14, 2009, equipment drain radioactive tank 5 was completely drained which created an unposted high radiation area. Two workers near the tank area received dose rate alarms indicating that unexpected radiological conditions existed. Radiation protection personnel responded to the area, performed surveys, and found an unexpected high radiation area outside of the posted high radiation area boundary. The highest dose rate outside the existing boundary was approximately 200 millirem/hour. The licensee entered this item into their corrective action program as Action Request 195295.

The failure to barricade and conspicuously post a high radiation area is a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, failure to post a high radiation area impacted the ability to adequately protect workers health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it was not an as low as is reasonably achievable finding, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. Additionally, this finding had human performance crosscutting aspects associated with work control in that the work planning did not appropriately plan work activities by incorporating risk insights and radiological safety H.3(a)(Section 2OS1).

Licensee-Identified Violations

None

REPORT DETAILS

Summary of Plant Status

The inspection period began with Columbia Generating Station operating at 100 percent power.

On April 2, 2009, the station began to coast down in power in preparation for Refueling Outage 19. On April 10, the station reduced power from 96 percent power to 85 percent power for economic dispatch and returned to 95 percent power on April 13. On April 16, the station reduced power from 94 percent power to 65 percent power to perform work on reactor feed water drive turbine 1B due to high bearing vibrations. Following this work, the station returned to 95 percent power on April 21, 2009. On May 8, the reactor was subsequently shutdown following a manual reactor scram from 88 percent reactor power and entered Forced Outage 09-02 due to maintenance leading to a loss of seal oil pressure and a subsequent loss of hydrogen in the main generator. The station then entered Refueling Outage 19 on May 9, 2009. The station completed Refueling Outage 19 on June 24, 2009 following reactor startup and closing of the main generator output breaker. On June 26, 2009, the station reached 75 percent power, as part of the planned power ascension, and initiated a manual reactor scram following a fire near the main turbine. The station entered Forced Outage 09-03 and remained in this outage until the end of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

1R01 Adverse Weather Protection

Summer Readiness for Offsite and Alternate ac Power

a. Inspection Scope

The inspectors performed a review of the licensees preparations for summer weather for selected systems, including conditions that could lead to loss-of-offsite power and conditions that could result from high temperatures. The inspectors reviewed the licensees procedures affecting these areas and the communications protocols between the transmission system operator and the plant to verify that the appropriate information was being exchanged when issues arose that could affect the offsite power system.

Examples of aspects considered in the inspectors review included:

• The coordination between the transmission system operator and the plant during off-normal or emergency events

• The explanations for the events

• The estimates of when the offsite power system would be returned to a normal state

• The notifications from the transmission system operator to the plant when the offsite power system was returned to normal During the inspection, the inspectors focused on plant-specific design features and the licensees procedures used to mitigate or respond to adverse weather conditions.

Additionally, the inspectors reviewed the Final Safety Analysis Report and performance requirements for systems selected for inspection, and verified that operator actions were appropriate as specified by plant-specific procedures. Specific documents reviewed during this inspection are listed in the attachment. The inspectors also reviewed corrective action program items to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program in accordance with station corrective action procedures. The inspectors reviews focused specifically on the following plant systems:

• Degraded Offsite Power Grid

• Loss of Offsite Power

These activities constitute completion of one readiness for summer weather affect on offsite and alternate ac power sample as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignments

Partial Walkdown

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant systems:

• High pressure core spray pump, April 5, 2009

• Diesel generator 3, following monthly surveillance testing, April 22, 2009

• Service water system B, June 3, 2009

The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors attempted to identify any discrepancies that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, Final Safety Analysis Report, technical specification requirements, administrative technical specifications, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The

inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three partial system walkdown samples as defined in Inspection Procedure 71111.04-05.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Quarterly Fire Inspection Tours

a. Inspection Scope

The inspectors conducted fire protection walkdowns that were focused on availability, accessibility, and the condition of firefighting equipment in the following risk-significant plant areas:

• Fire area RC-2, cable spreading room, March 31, 2009

• Fire area RC-14/1, Division 1 switchgear room, April 21, 2009

• Fire door W-DOOR-C216, April 22, 2009

• Fire area TG-7, hydrogen seal oil room, May 8, 2009

• Fire area R-1/1, reactor 522' NW quadrant, May 26, 2009

The inspectors reviewed areas to assess if licensee personnel had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant; effectively maintained fire detection and suppression capability; maintained passive fire protection features in good material condition; and had implemented adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features, in accordance with the licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plants Individual Plant Examination of External Events with later additional insights, their potential to affect equipment that could initiate or mitigate a plant transient, or their impact on the plants ability to respond to a security event. Using the documents listed in the attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed, that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified

during the inspection were entered into the licensees corrective action program.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of five quarterly fire-protection inspection samples as defined in Inspection Procedure 71111.05AQ-05.

b. Findings

No findings of significance were identified.

.2 Annual Fire Protection Drill Observation

a. Inspection Scope

On April 16, 2009, the inspectors observed fire brigade activation for an unannounced backshift fire drill. The observation evaluated the readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee staff identified deficiencies, openly discussed them in a self-critical manner at the drill debrief, and took appropriate corrective actions. Specific attributes evaluated were:

(1) proper wearing of turnout gear and self-contained breathing apparatus;

(2) proper use and layout of fire hoses;

(3) employment of appropriate fire fighting techniques;

(4) sufficient firefighting equipment brought to the scene;

(5) effectiveness of fire brigade leader communications, command, and control;

(6) search for victims and propagation of the fire into other plant areas;

(7) smoke removal operations;

(8) utilization of pre planned strategies;

(9) adherence to the preplanned drill scenario; and

(10) drill objectives.

These activities constitute completion of one annual fire-protection inspection sample as defined in Inspection Procedure 71111.05AQ-05.

b. Findings

No findings of significance were identified.

1R08 In-service Inspection Activities

From May 18-22, 2009, the inspectors performed Inspection Procedure 71111.08, Inservice Inspection Activities. Inspection Procedure 71111.08 requires a minimum sample size, for boiling water reactors, of one for Section 02.01. The inspectors fulfilled the requirements of Inspection Procedure 71111.08-05.

.1 Inspection Activities Other Than Steam Generator Tube Inspection, Pressurized Water

Reactor Vessel Upper Head Penetration Inspections, Boric Acid Corrosion Control (71111.08-02.01)

a. Inspection Scope

This inspection assesses the effectiveness of the licensees program for monitoring degradation of vital system boundaries. The inspection includes a review of the licensees nondestructive examination and welding programs. The inspectors are to verify that inservice inspection and welding activities are performed in accordance with ASME Code, other regulatory requirements, and licensee commitments.

The inspectors reviewed three volumetric examinations and 12 surface examinations.

From these 15 examinations, the inspectors observed three ultrasonic examinations:

The inspectors directly observed the following examinations:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE

Main Feedwater Nozzle N4-B; weld AB-9 UT Main Feedwater Nozzle N4-B: weld AB-10 UT Reactor Pressure Vessel Nozzle N4-C; Nozzle to vessel weld UT Reactor Pressure Vessel Feedwater Sparger Pin at 305 degree location FSP-305 VT-1 Reactor Pressure Vessel Feedwater Sparger Bracket at 295 deg. location FSB-295 VT-1 Reactor Pressure Vessel Jet Pump Set Screw Shroud Side JP AS-1 SS VT-1

The inspectors reviewed records for the following nondestructive examinations:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE 6-07-1-77 MS Weld XI-2-1 PT 6-07-1-78 MS Weld XI-1-1 PT 6-07-1-75 MS Weld XI-1-1, XI-2-1 PT 3RHM-006 RHR Heat Exchanger Support Welds MT

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE 3MSM-004 MS Welded Lugs MT 3RHM-002 RHR-P-2BN-3, -2BC-4, and -

2BC-5 MT 3RHM-003 RHR-P-2BN-1 MT 3RHM-004 RHR Inlet Nozzle/Top Head MT 3RHM-005 RHR outlet nozzle to shell weld MT

During the review and observation of each examination, the inspectors verified that activities were performed in accordance with ASME Boiler and Pressure Vessel Code requirements and applicable procedures. Indications were compared with previous examinations and dispositioned in accordance with ASME Code and approved procedures. The qualifications of all nondestructive examination technicians performing the inspections were verified to be current.

None of the above observed or reviewed nondestructive examinations identified any relevant indications and cognizant-licensee personnel stated that no relevant indications were accepted by the licensee for continued service.

The inspectors reviewed the following volumetric examinations from the previous outage, which identified relevant indications that were analytically evaluated and accepted by the licensee for continued service:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Reactor Pressure Vessel RPV Head Weld AH UT RRC 4RRC(4)B-11 UT

The licensees acceptance was in accordance with ASME Code or an NRC approved alternative, and the indications were examined for acceptability of continued service.

The inspectors examined the following welding that was performed on pressure boundary, risk significant systems:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Reactor Pressure Vessel RPV Head Weld AH UT RRC 4RRC(4)B-11 UT

The inspectors verified, by review, that the welding procedure specifications and the welders had been properly qualified in accordance with ASME Code,Section IX, requirements. The inspectors also verified, through observation and record review, that essential variables for the welding process were identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specifications. Specific documents reviewed during this inspection are listed in the attachment.

b. Findings

No findings of significance were identified.

.2 Identification and Resolution of Problems (71111.08-02.05)

a. Inspection Scope

The inspection procedure requires review of a sample of problems associated with inservice inspections documented by the licensee in the corrective action program for appropriateness of the corrective actions.

The inspectors reviewed 10 condition reports which dealt with inservice inspection activities and found the corrective actions were appropriate. The specific condition reports reviewed are listed in the documents reviewed section. From this review the inspectors concluded that the licensee has an appropriate threshold for entering issues into the corrective action program and has procedures that direct a root cause evaluation when necessary. The licensee also has an effective program for applying industry operating experience. Specific documents reviewed during this inspection are listed in the attachment.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program

a. Inspection Scope

On April 6, 2009, the inspectors observed a crew of licensed operators in the plants simulator to verify that operator performance was adequate, evaluators were identifying and documenting crew performance problems, and training was being conducted in accordance with licensee procedures. The training scenario involved a seismic event that caused a loss of reactor pressure vessel level indication that required reactor pressure vessel flooding to ensure adequate core cooling. The inspectors evaluated the following areas:

• Licensed operator performance

• Crews clarity and formality of communications

• Crews ability to take timely actions in the conservative direction

• Crews prioritization, interpretation, and verification of annunciator alarms

• Crews correct use and implementation of abnormal and emergency procedures

• Control board manipulations

• Oversight and direction from supervisors

• Crews ability to identify and implement appropriate technical specification actions and emergency plan actions and notifications

The inspectors compared the crews performance in these areas to pre-established operator action expectations and successful critical task completion requirements.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one quarterly licensed-operator requalification program sample as defined in Inspection Procedure 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following risk significant systems:

• Work Order 01160103, Replace seal drywell hatch, April 1, 2009

• Work Order 01141820, Replace SW-V-107, crane work around service water ponds, June 15, 2009

The inspectors reviewed events such as where ineffective equipment maintenance has resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

• Implementing appropriate work practices

• Identifying and addressing common cause failures

• Scoping of systems in accordance with 10 CFR 50.65(b)

• Characterizing system reliability issues for performance

• Charging unavailability for performance

• Trending key parameters for condition monitoring

• Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or (a)(2)

• Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as requiring the establishment of appropriate and adequate goals and corrective actions for systems classified as not having adequate performance, as described in 10 CFR 50.65(a)(1)

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of two quarterly maintenance effectiveness samples as defined in Inspection Procedure 71111.12-05.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors reviewed licensee personnel's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk-significant and safety-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

• Work Order 01162218, heavy lift of REA-Fan-1B, March 31, 2009

• Test of reactor core isolation coolant with work on backup transformer, April 9, 2009

• Work Order 01138221, lower spent fuel pool level, April 20, 2009

• Tour of temporary scaffolding installed in the reactor building in preparation for Refueling Outage 19, April 29, 2009

• Action Report/Condition Report 198282, evaluation of 70 ton Grove crane, May 29, 2009

• Action Report/Condition Report 198269, 70 ton Grove crane, May 30, 2009

• Action Report/Condition Report 198607, evaluate generic operation of 70 ton Grove crane and boom truck, June 2, 2009

• Work Order 01141820, rebuild service water valve SW-V-170B, June 15, 2009 The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4)and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of eight maintenance risk assessments and emergent work control inspection samples as defined in Inspection Procedure 71111.13-05.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• Action Report/Condition Report 194990, reactor building siding liner panels not continuous, April 9, 2009

• Action Report/Condition Report 194206, reactor building ventilation, April 27, 2009

• Action Report/Condition Report 196046, incomplete operability evaluation for Action Report/Condition Report 191584, and Action Report/Condition Report 191584, CCH-RV-2A is leaking by at 150 drops per minute, April 29, 2009

• Action Report/Condition Report 190083, RCIC-LS-10 shaking, April 29, 2009 The inspectors selected these potential operability issues based on the risk-significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that technical specification operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the technical specifications and Final Safety Analysis Report to the licensees evaluations, to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four operability evaluations inspection samples as defined in Inspection Procedure 71111.15-04

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

.1 Temporary Plant Modifications

a. Inspection Scope

The inspectors reviewed the following temporary modification to verify that the safety functions of important safety systems were not degraded:

• Temporary Modification Request 09-007, fuel pool cooling motor 1A, April 22, 2009

The inspectors reviewed the temporary modification and the associated safety evaluation screening against the system design bases documentation, including the Final Safety Analysis Report and the technical specifications, and verified that the modification did not adversely affect the system operability/availability. The inspectors also verified that the installation and restoration were consistent with the modification documents and that configuration control was adequate. Additionally, the inspectors

verified that the temporary modification was identified on control room drawings, appropriate tags were placed on the affected equipment, and licensee personnel evaluated the combined effects on mitigating systems and the integrity of radiological barriers.

These activities constitute completion of one sample for temporary plant modifications as defined in Inspection Procedure 71111.18-05.

b. Findings

No findings of significance were identified.

.2 Permanent Plant Modifications

a. Inspection Scope

The inspectors reviewed key affected parameters associated with energy needs, materials/replacement components, timing, heat removal, control signals, equipment protection from hazards, operations, flow paths, pressure boundary, ventilation boundary, structural, process medium properties, licensing basis, and failure modes for the permanent modification listed below:

• Work Order 01172331, RRC-V-23A Repair, June 16, 2009

The inspectors verified that modification preparation, staging, and implementation did not impair emergency/abnormal operating procedure actions, key safety functions, or operator response to loss of key safety functions; post modification testing will maintain the plant in a safe configuration during testing by verifying that unintended system interactions will not occur, systems, structures and components performance characteristics still meet the design basis, the appropriateness of modification design assumptions, and the modification test acceptance criteria will be met; and licensee personnel identified and implemented appropriate corrective actions associated with permanent plant modifications. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one sample for permanent plant modifications as defined in Inspection Procedure 71111.18-05.

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing

a. Inspection Scope

The inspectors reviewed the following postmaintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

• Work Order 01169535, postmaintenance testing, April 21, 2009

• Action Report/Condition Report 195935, reactor core isolation cooling turbine bearing oil level is slightly in yellow band, April 26, 2009

• RHR-V209 check valve repair and replacement parts, May 2009

• Work Order 01165944, OSP-RCIC/IST-Q701, reactor core isolation cooling operability test, June 22, 2009

• Main steam relief valve testing, June 23, 2009

• RRC-V-23A, leak test at 1000 pounds, June 23, 2009

• Work Order 01137764, inspection of RRC-V-24A gland packing leak-off at 1000 pounds, June 24, 2009

• Work Order 01139947, RFT-DT digital electro-hydraulic control system testing, June 25, 2009 The inspectors selected these activities based upon the structure, system, or component's ability to affect risk. The inspectors evaluated these activities for the following:

• The effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed

• Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate The inspectors evaluated the activities against the technical specifications, the Final Safety Analysis Report, 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with postmaintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of 8 postmaintenance testing inspection sample(s)as defined in Inspection Procedure 71111.19-05.

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors reviewed the outage safety plan and contingency plans for the Columbia Generating Station Unit 2 refueling outage, conducted May 9 to June 24, 2009, to confirm that licensee personnel had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. During the refueling outage, the inspectors observed portions of the shutdown and cooldown processes and monitored licensee controls over the outage activities listed below.

• Configuration management, including maintenance of defense-in-depth, is commensurate with the outage safety plan for key safety functions and compliance with the applicable technical specifications when taking equipment out of service.

• Clearance activities, including confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing.

• Installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error.

• Status and configuration of electrical systems to ensure that technical specifications and outage safety-plan requirements were met, and controls over switchyard activities.

• Monitoring of decay heat removal processes, systems, and components.

• Verification that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system.

• Reactor water inventory controls, including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss.

• Controls over activities that could affect reactivity.

• Maintenance of secondary containment as required by the technical specifications.

• Refueling activities, including fuel handling and sipping to detect fuel assembly leakage.

• Startup and ascension to full power operation, tracking of startup prerequisites, walkdown of the drywell (primary containment) to verify that debris had not been left which could block emergency core cooling system suction strainers, and reactor physics testing.

• Licensee identification and resolution of problems related to refueling outage activities.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one refueling outage and other outage inspection sample as defined in Inspection Procedure 71111.20-05.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the Final Safety Analysis Report, procedure requirements, and technical specifications to ensure that the surveillance activities listed below demonstrated that the systems, structures, and/or components tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the significant surveillance test attributes were adequate to address the following:

• Preconditioning

• Evaluation of testing impact on the plant

• Acceptance criteria

• Test equipment

• Procedures

• Jumper/lifted lead controls

• Test data

• Testing frequency and method demonstrated technical specification operability

• Test equipment removal

• Restoration of plant systems

• Fulfillment of ASME Code requirements

• Updating of performance indicator data

• Engineering evaluations, root causes, and bases for returning tested systems, structures, and components not meeting the test acceptance criteria were correct

• Reference setting data

• Annunciators and alarms setpoints

The inspectors also verified that licensee personnel identified and implemented any needed corrective actions associated with the surveillance testing.

• Work Order 01162402, stack monitor monthly source and channel check, April 1, 2009

• Work Order 01161145, TSP-DG2-X501 diesel generator shutdown logic checks, April 2, 2009

• CSP-SLC-M101, standby liquid control boron, April 7, 2009

• Valve inservice test - Work Order 01147789, RHR-V-209 repair and partial OSP, April 7, 2009

• Test of reactor core isolation cooling with work on backup transformer, April 9, 2009

• Work Order 01143605, E-CB01/7 H1 TOC switch replacement, April 14, 2009

• Containment isolation valve local leak rate test - Work Order 01138221, inspect FPC-V-107, April 20, 2009

• Reactor coolant system leakage detection surveillance - Work Order 01164057, ISP-LD-S901, leak detection monitor 1A surveillance, May 19, 2009 Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of eight surveillance testing inspection samples as defined in Inspection Procedure 71111.22-05.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational and Public Radiation Safety

2OS1 Access Control to Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess licensee performance in implementing physical and administrative controls for airborne radioactivity areas, radiation areas, high radiation areas, and worker adherence to these controls. The inspector used the requirements in 10 CFR Part 20, the technical specifications, and the licensees procedures required by technical specifications as criteria for determining compliance. During the inspection, the inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors performed independent radiation dose rate measurements and reviewed the following items:

• Performance indicator events and associated documentation packages reported by the licensee in the Occupational Radiation Safety Cornerstone

• Controls (surveys, posting, and barricades) of radiation, high radiation, or airborne radioactivity areas

• Radiation work permits, procedures, engineering controls, and air sampler locations

• Conformity of electronic personal dosimeter alarm set points with survey indications and plant policy; workers knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarms

• Barrier integrity and performance of engineering controls in airborne radioactivity areas

• Adequacy of the licensees internal dose assessment for any actual internal exposure greater than 50 millirem committed effective dose equivalent

• Physical and programmatic controls for highly activated or contaminated materials (non-fuel) stored within spent fuel and other storage pools

• Self-assessments, audits, licensee event reports, and special reports related to the access control program since the last inspection

• Corrective action documents related to access controls

• Licensee actions in cases of repetitive deficiencies or significant individual deficiencies

• Radiation work permit briefings and worker instructions

• Adequacy of radiological controls, such as required surveys, radiation protection job coverage, and contamination control during job performance

• Dosimetry placement in high radiation work areas with significant dose rate gradients

• Changes in licensee procedural controls of high dose rate - high radiation areas and very high radiation areas

• Controls for special areas that have the potential to become very high radiation areas during certain plant operations

• Posting and locking of entrances to all accessible high dose rate - high radiation areas and very high radiation areas

• Radiation worker and radiation protection technician performance with respect to radiation protection work requirements Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of 21 of the required 21 samples as defined in Inspection Procedure 71121.01-05.

b. Findings

Introduction:

The inspectors reviewed a Green, self-revealing noncited violation of Technical Specification 5.7.1 for failure to barricade and conspicuously post a high radiation area.

Description:

On April 14, 2009, equipment drain radioactive tank 5 was partially drained in support of cleaning the tank. At approximately one percent capacity, radiation protection personnel surveyed the tank area to verify the dose rates at the existing high radiation area boundary. The highest dose rate at the boundary was approximately 60 millirem per hour. Operations personnel continued with the planned tagout of the tank without radiation protection personnel being aware that the tank would be completely drained of water. Approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the draining of the tank, two workers near the tank area received dose rate alarms indicating that unexpected radiological conditions existed. Radiation protection personnel responded to the area, performed surveys, and found an unexpected high radiation area outside of the posted high radiation area boundary. The highest dose rate outside the existing boundary was approximately 200 millirem per hour. Therefore, the high radiation area boundary had to be extended out another 4 feet. Dose rates in the general work area of the tank were 50-60 millirem per hour as opposed to 5-8 millirem per hour prior to draining the tank.

Analysis:

The failure to barricade and conspicuously post a high radiation area is a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and affected the cornerstone objective, in that, failure to post a high radiation area impacted the ability to adequately protect workers health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it was not an ALARA finding, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The finding was self-revealing because the licensee was alerted to the situation when the workers received unexpected electronic dosimeter alarms.

Additionally, this finding had human performance crosscutting aspects associated with work control in that work planning did not appropriately plan work activities by incorporating risk insights and radiological safety H.3(a).

Enforcement:

Technical Specification 5.7.1 requires, in part, that for high radiation areas with dose rates not exceeding 1.0 rem per hour (at 30 centimeters from the radiation sources or from any surface penetrated by the radiation), each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Pursuant to 10 CFR 20.1003, a high radiation area means an area, accessible to individuals, in which radiation levels from radiation sources external to the body could result in an individual receiving a dose equivalent in excess of 0.1 rem (1 mSv) in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 centimeters from the radiation source or 30 centimeters from any surface that the radiation penetrates. Contrary to the above, on April 14, 2009, the licensee failed to barricade and conspicuously post a high radiation area, with dose rates as high as 200 millirem per hour. Specifically, equipment drain radioactive tank 5 was completely drained creating an unposted high radiation area. Because this failure to barricade and conspicuously post a high radiation area is of very low safety significance and has been entered into the licensees corrective action program as action request 195295, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000397/2009003-01, Failure to Barricade and Conspicuously Post a High Radiation Area.

2OS2 ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspector assessed licensee performance with respect to maintaining individual and collective radiation exposures ALARA. The inspector used the requirements in 10 CFR 20 and the licensees procedures required by technical specifications as criteria for determining compliance. The inspector interviewed licensee personnel and reviewed the following:

• Interfaces between operations, radiation protection, maintenance, maintenance planning, scheduling and engineering groups

• Integration of ALARA requirements into work procedure and radiation work permit (or radiation exposure permit) documents

• Dose rate reduction activities in work planning

• Use of engineering controls to achieve dose reductions and dose reduction benefits afforded by shielding

• Workers use of the low dose waiting areas

• Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of 2 of the required 15 samples and 4 of the optional samples as defined in Inspection Procedure 71121.02-05.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. Inspection Scope

The inspectors performed a review of the performance indicator data submitted by the licensee for the first quarter of 2009 for any obvious inconsistencies prior to its public release in accordance with Inspection Manual Chapter 0608, Performance Indicator Program.

This review was performed as part of the inspectors normal plant status activities and, as such, did not constitute a separate inspection sample.

b. Findings

No findings of significance were identified.

.2 Safety System Functional Failures (MS05)

a. Inspection Scope

The inspectors sampled licensee submittals for the Safety System Functional Failures performance indicator for the period from the first quarter 2008 through the first quarter

2009. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, and NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73 definitions and guidance were used. The inspectors reviewed the licensees operator narrative logs, operability assessments, maintenance rule records, maintenance work orders, issue reports, event reports and NRC integrated inspection reports for the period of January 2008 through March 2009 to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one safety system functional failures sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.3 Reactor Coolant System Specific Activity (BI01)

a. Inspection Scope

The inspectors sampled licensee submittals for the Reactor Coolant System Specific Activity performance indicator for the period from the first quarter 2008 through first quarter 2009. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees reactor coolant system chemistry samples, technical specification requirements, issue reports, event reports and NRC integrated inspection reports for the period of January 2008 through March 2009 to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. In addition to record reviews, the inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one reactor coolant system specific activity sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.4 Reactor Coolant System Leakage (BI02)

a. Inspection Scope

The inspectors sampled licensee submittals for the Reactor Coolant System Leakage performance indicator for the period from the first quarter 2008 through first quarter 2009. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees operator logs, reactor coolant system leakage tracking data, issue reports, event reports and NRC integrated inspection reports for the period of January 2008 through March 2009 to validate the accuracy of the submittals. The inspectors also reviewed the licensees issue report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

These activities constitute completion of one reactor coolant system leakage sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.5 Occupational Exposure Control Effectiveness (OR01)

a. Inspection Scope

The inspectors sampled licensee submittals for the Occupational Radiological Occurrences performance indicator for the fourth quarter of 2008 and first quarter of 2009. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees assessment of the performance indicator for occupational radiation safety to determine if indicator related data was adequately assessed and reported. To assess the adequacy of the licensees performance indicator data collection and analyses, the inspectors discussed with radiation protection staff, the scope and breadth of its data review, and the results of those reviews. The inspectors independently reviewed electronic dosimetry dose rate and accumulated dose alarm and dose reports and the dose assignments for any intakes that occurred during the time period reviewed to determine if there were potentially unrecognized occurrences. The inspectors also conducted walkdowns of numerous locked high and very high radiation area entrances to determine the adequacy of the controls in place for these areas.

These activities constitute completion of the occupational radiological occurrences sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

.6 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual

Radiological Effluent Occurrences (PR01)

a. Inspection Scope

The inspectors sampled licensee submittals for the Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences performance indicator for the fourth quarter of 2008 and first quarter of 2009. To determine the accuracy of the performance indicator data reported during those periods, performance indicator definitions and guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The inspectors reviewed the licensees issue report database and selected individual reports generated since this indicator was last reviewed to identify any potential occurrences such as unmonitored, uncontrolled, or improperly calculated effluent releases that may have impacted offsite dose. The inspectors reviewed gaseous effluent summary data and the results of associated offsite dose calculations for selected dates during the fourth quarter of 2008 to determine if indicator results were accurately reported. The inspectors also reviewed the licensees methods for quantifying gaseous and liquid effluents and determining effluent dose. Additionally, the inspectors reviewed the licensees historical 10 CFR Part 50.75(g) file and selectively reviewed the licensees analysis for discharge pathways resulting from a spill, leak, or unexpected liquid discharge focusing on those incidents which occurred over the last few years.

These activities constitute completion of the radiological effluent technical specifications/offsite dose calculation manual radiological effluent occurrences sample as defined in Inspection Procedure 71151-05.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities

and plant status reviews to verify that they were being entered into the licensees corrective action program at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. The inspectors reviewed attributes that included: the complete and accurate identification of the problem; the timely correction, commensurate with the safety significance; the evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, and previous occurrences reviews; and the classification, prioritization, focus, and timeliness of corrective actions. Minor issues entered into the licensees corrective action program because of the inspectors observations are included in the attached list of documents reviewed.

These routine reviews for the identification and resolution of problems did not constitute any additional inspection samples. Instead, by procedure, they were considered an integral part of the inspections performed during the quarter and documented in Section 1 of this report.

b. Findings

Introduction.

The inspectors reviewed a Green self-revealing finding for the failure of Energy Northwest to implement the standards and guidance provided in Site Wide Procedure SWP-CAP-01, Corrective Action Program, Revision 17. As a result, the licensee failed to implement interim corrective actions, for previously identified single point vulnerability in the main generator seal oil system, which resulted in a reactor scram while performing testing on the system.

Description.

On May 8, 2009, during scheduled pre-outage generator seal oil system back-up testing, the generator air side seal oil system unexpectedly lost pressure due to an abruptly plugged Cuno filter, SO-F-1. This resulted in a loss of seal oil, which in turn caused a failure of the hydrogen seal to maintain generator gas pressure and hydrogen escaped through the shaft seals. Due to the release of hydrogen gas from the generator, generator pressure dropped from 72 psi to 30 psi, the station entered Procedure ABN-GENERATOR, Main Generator Trouble, Revision 4, lowered power from 1000 MW to 925 MW, and then manually scrammed the reactor.

The Cuno filter plugging during maintenance is a known problem that was first identified in September 1989 as documented in PER 289-0544. An upgrade to the filtering system was approved in November 1990 but was subsequently cancelled in October 1996 due to budget constraints. Action Request 1485 was initiated in September 2000 recommending upgrading the system to a high efficiency duplex filter assembly. Several requests were made and Action Request 1485 was approved in March 2007 to be implemented in fiscal year 2009. An administrative error, as documented in Action Request 1485-4 on March 11, 2008, scheduled the filter upgrade for fiscal year 2010 or 2011. A known corrective action to prevent the Cuno filter from plugging is to have an operator manually rotate the filter.

Analysis.

Energy Northwests failure to implement the standards and guidance provided in Procedure SWP-CAP-01 is a performance deficiency. Specifically, delaying the corrective action of a single-point vulnerability and not implementing any interim corrective actions resulted in a plant scram. Procedure SW-CAP-01, Section 4.11.1.g states in part that, Long term corrective actions are assigned a due date commensurate with the safety significance of the condition provided reasonable efforts are made to complete the corrective actions promptly or at the first available opportunity unless appropriate justification is provided for a longer completion schedule. Section 4.11.1.f of Procedure SWP-CAP-01 states in part that, If the actions in a CAP cannot be implemented in a timely manner, the plan should include interim actions. Contrary to this procedure, inspectors found that timely action had not been taken to address a single-point vulnerability in the seal oil system.

The inspectors used NRC Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, to determine that the finding was more than minor because it was an equipment performance issue that affected the Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609.4, Phase 1 - Initial Screening and Characterization of Findings, this finding was determined to be of very low safety significance (Green)because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The finding had a crosscutting aspect in the area of problem identification and resolution associated with operating experience [P.2.(b)], because the licensee failed to implement operational experience through changes to station processes, procedures, equipment, and training programs (Section 4OA2).

Enforcement.

Because this finding does not involve a violation of regulatory requirements and has very low safety significance, it is identified as:

FIN 05000397/2009003-02, Reactor Scram Due to Seal Oil Leak for Main Generator.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program. The inspectors accomplished this through review of the stations daily corrective action documents.

The inspectors performed these daily reviews as part of their daily plant status monitoring activities and, as such, did not constitute any separate inspection samples.

b. Findings

No findings of significance were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensees corrective action program and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors focused their review on repetitive equipment issues, but also considered the results of daily corrective action item screening discussed in Section 4OA2.2, above, licensee trending efforts, and licensee human performance results. The inspectors nominally considered the 6-month period of January through June 2009, although some examples expanded beyond those dates where the scope of the trend warranted.

The inspectors also included issues documented outside the normal corrective action program in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance audit/surveillance reports, self-assessment reports, and maintenance rule assessments.

The inspectors compared and contrasted their results with the results contained in the licensees corrective action program trending reports. Corrective actions associated with a sample of the issues identified in the licensees trending reports were reviewed for adequacy.

These activities constitute completion of one single semi-annual trend inspection sample as defined in Inspection Procedure 71152-05.

b. Findings

No findings of significance were identified.

4OA3 Event Follow-up

.1 May 8, 2009, Manual Reactor Scram

a. Inspection Scope

On May 8, 2009, the inspectors observed and evaluated Energy Northwests response to a scram while the reactor was operating at 87 percent power. Specifically, Energy Northwests failure to correct a known single-point vulnerability caused a loss of seal oil pressure and a resultant plant scram. The inspectors responded to the site and verified plant conditions by observing key plant parameters, annunciator status, and observing the current status of safety related mitigating equipment to ensure that the plant was stable. The inspectors also observed reactor operator actions in response to the reactor scram and senior reactor operators evaluation of plant conditions and oversight of the reactor operators to ensure that operators were adhering to plant procedures. The inspectors also reviewed Energy Northwests evaluation of the root cause of the scram.

b. Findings

One finding was identified, as documented in Section 4OA2.

.2 (Closed) Licensee Event Report 05000397/2009-001-00:

Reactor Scram Due to Turbine Control System Trip Header Depressurization This Licensee Event Report documents the automatic scram that occurred on February 8, 2009, due to a pressure transient following on-line maintenance of the digital electro-hydraulic system. Energy Northwest determined the cause of the event to be design deficiencies in the on-line serviceable assembly, which allowed system conditions that resulted in a reactor trip. See Inspection Report 05000397/2009002 for a discussion of a self-revealing finding associated with this issue. The inspectors completed a review of the Licensee Event Report and did not identify any other violations of regulatory requirements or findings. This Licensee Event Report is closed.

Specific documents reviewed are described in the attachment to this report.

.3 June 26, 2009 Manual Reactor Scram

a. Inspection Scope

On June 26, 2009, the inspectors observed and evaluated Energy Northwests response to a scram while the reactor was operating at 65 percent power. The inspectors responded to the site and verified plant conditions by observing key plant parameters, annunciator status, and observing the current status of safety related mitigating equipment to ensure that the plant was stable. The inspectors also observed reactor operator actions in response to the reactor scram and senior reactor operators evaluation of plant conditions and oversight of the reactor operators to ensure that operators were adhering to plant procedures.

b. Findings

No findings of significance were identified. The inspectors will review the licensees root cause analysis and corrective actions.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors performed observations of security force personnel and activities to ensure that the activities were consistent with Columbia Generating Stations security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours. Specific documents reviewed are described in the attachment to this report.

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 Institute of Nuclear Power Operations (INPO) Plant Assessment Report Review

a. Inspection Scope

On May 5, 2009, the inspectors completed a review of the final report for the INPO plant assessment of Columbia Generating Station conducted in October 2008. The inspectors reviewed the report to ensure that issues identified were consistent with the NRC perspectives of licensee performance and to verify if any significant safety issues were identified that required further NRC follow-up.

b. Findings

No findings of significance were identified

4OA6 Meetings

Exit Meeting Summary

On May 21, 2009, the inspectors presented the results of this inservice inspection to Mr. S. Gambir, Vice President of Technical Services, and other members of licensee management. Licensee management acknowledged the inspection findings. The inspectors also confirmed that no proprietary material was reviewed during the inspection.

On May 22, 2009, the inspectors presented the occupational and public radiation safety inspection results to Mr. S. Oxenford, Vice President Nuclear Generation, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On July 7, 2009, the inspectors presented the inspection results to Mr. S. Oxenford, Vice President, Nuclear Generation, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On July 13, 2009, the senior resident inspector conducted a final exit meeting with Mr. M. Humphreys, Licensing Supervisor, and other members of the licensing staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Atkinson, Vice President, Operations Support

G. Cullen, Manager, Regulatory Programs

M. Davis, Manager, Radiation Protection

J. Frisco, General Manager, Engineering

S. Gambhir, Vice President, Technical Services

R. Garcia, Specialist, Licensing

W. LaFramboise, System Engineering Manager

T. Lynch, Plant General Manager

J. Parrish, Chief Executive Officer

F. Schill, Licensing

M. Shepherd, Supervisor, Radiation Protection

C. Tiemans, Supervisor, Radiation Protection

C. Whitcomb, Vice President, Organizational Performance and Staffing

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Opened and Closed

05000397/2009003-01 NCV Failure to Barricade and Conspicuously Post a High Radiation Area (Section 4OA2)

05000397/2009003-02 FIN Reactor Scram Due to Seal Oil Leak for Main Generator (Section 2OS1)

Closed

05000397/2009-001-00 LER Reactor Scram Due to Turbine Control System Trip Header Depressurization (Section 4OA3)