IR 05000397/2005008

IR 05000397-05-008,on 3/14/05 - 4/27/05; Columbia Generating Station; Biennial Baseline Inspection of the Identification and Resolution of Problems

ML051610699
Person / Time
Site:	Columbia
Issue date:	06/10/2005
From:	Laura Smith Division of Reactor Safety IV
To:	Parrish J Energy Northwest
References
IR-05-008
Download: ML051610699 (28)

Text

une 10, 2005

SUBJECT:

COLUMBIA GENERATING STATION - NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000397/2005008

Dear Mr. Parrish:

On April 27, 2005, the U. S. Nuclear Regulatory Commission (NRC) completed a team inspection at your Columbia Generating Station. The enclosed report documents the inspection findings, which were debriefed on March 31, 2005, with you and other members of your staff.

Following additional in-office inspection, a telephonic exit meeting was conducted on April 27, 2005 with Mr. Roberto Torres, Quality Assurance and Corrective Action Manager, and other members of your staff.

This inspection examined activities conducted under your license as they relate to the identification and resolution of problems, compliance with the Commissions rules and regulations and with the conditions of your license. The team reviewed approximately 370 condition reports and problem evaluation requests. In addition, the team examined other selected documents, observed activities, and interviewed personnel in support of the inspection effort.

On the basis of the sample selected for review, the team concluded that, in general, your processes to identify, prioritize, evaluate, and correct problems were effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. Notwithstanding the above, the team observed that, for the past several years, poor evaluations of breaker and switchgear problems still challenge the site, as NRC identified and self-revealing issues continue to surface. A similar observation was included in the last problem identification and resolution inspection report. The team concluded that a positive safety-conscious work environment existed at Columbia Generating Station.

The report documents three findings that were evaluated under the Significance Determination Process as having very low safety significance (Green). The NRC determined that two violations were associated with these findings. The violations are being treated as non-cited violations because they are of very low safety significance and because they have been entered into your corrective action program consistent with Section VI.A of the Enforcement Policy. If you contest the violations or the significance of these non-cited violations, you should provide a

Energy Northwest -2-response within 30 days of the date of the inspection report, with the basis for your denial, to the U.S. Nuclear Regulator Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011;the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Columbia Generating Station.

In accordance with 10 CFR 2.390 of the NRCs 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 (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

//RA//

Linda Joy Smith, Chief Engineering Branch 2 Division of Reactor Safety Docket: 50-397 License: NPF-21

Enclosure:

NRC Inspection Report 05000397/2005008

REGION IV==

Docket: 50-397 License: NPF-21 Report: 05000397/2005008 Licensee: Energy Northwest Facility: Columbia Generating Station Location: Richland, Washington Dates: March 14 through April 27, 2005 Inspectors: T. Jackson, Senior Resident Inspector - Team Leader Z. Dunham, Senior Resident Inspector G. Werner, Senior Project Engineer T. McKernon, Senior Operations Engineer D. Holman, Security Inspector Approved By: Linda Joy Smith, Chief Engineering Branch 2 Division of Reactor Safety ATTACHMENT: Supplemental Information Enclosure

SUMMARY OF FINDINGS

IR 05000397/2005008; 3/14/05 - 4/27/05; Columbia Generating Station; biennial baseline inspection of the identification and resolution of problems. Findings were identified in the areas of Identification and Resolution; Prioritization and Evaluation; and Effectiveness of Corrective Actions.

The inspection was conducted by two senior resident inspectors, a senior operations engineer, a senior project engineer, and a security inspector. One self-revealing and one NRC-identified Green non-cited violation (NCV), one self-revealing Green finding, and two unresolved items were identified during this inspection. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 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 NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649,

Reactor Oversight Process, Revision 3, dated July 2000.

Identification and Resolution of Problems

• The team reviewed approximately 370 condition reports, apparent and root cause analyses, as well as other documents, to assess problem identification and resolution activities. While the licensees processes were generally effective, the team observed that, for approximately the last four years, poor electrical engineering evaluations of breaker and switchgear problems resulted in a disproportionate number of NRC identified and self-revealing issues. In addition, several of the findings were related to inadequate consideration of seismic requirements. A similar performance concern was documented in the last problem identification and resolution assessment.

The team concluded that a safety-conscience work environment existed at the Columbia Generating Station. The team determined that employees and contractors felt free to enter issues into the corrective action program and raise safety concerns to their supervision, to the employees concern program, and to the NRC. All the interviewees believed that potential safety issues were being addressed. However, the team received an isolated comment regarding receptiveness of some supervisors to initiating condition reports. Energy Northwest management planned to take corrective measures to address this comment.

Cornerstone: Mitigating Systems

Green.

The team identified a non-cited violation of 10 CFR 50, Appendix B,

Criterion XVI (Corrective Actions) for the failure to identify nonconforming breaker settings (conditions adverse to quality). The licensee had identified that overcurrent settings were incorrect for General Electric Type TEC molded-case circuit breakers but did not evaluate the potential for the same problem to occur with other molded case circuit breakers. In response to NRC questions, additional problems were identified. Two safety-related breakers and one non-

safety related breaker required recalibration to correctly establish the breaker trip points. The finding had crosscutting aspects associated with problem evaluation.

The failure to perform an adequate engineering evaluation of a condition adverse to quality was a performance deficiency. The finding was more than minor because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding had very low safety significance because it did not result in a loss of safety function, a loss of a safety-related train for greater than its Technical Specification allowed outage time, the loss of risk-significant non-Technical Specification trains for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event (Section 4OA2.e(2)iii).

Green.

The team documented a self-revealing finding to address an inadequate maintenance procedure. The procedure failed to specify the correct relay contact configuration for a startup transformer control circuit relay. When maintenance craftsmen installed the relay and placed the unit in service, the startup transformer tripped off-line. It was recovered approximately 20 minutes later. The issue had crosscutting aspects associated with human performance, adequacy of procedures.

The failure to provide an adequate procedure for startup transformer work was a performance deficiency. The finding was more than minor because it affected the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. The finding is similar to NRC Inspection Manual, Chapter 0612, Appendix E, non-minor Example 5.b, in that the equipment was returned to service before the problem was identified.

The finding was of very low safety significance because it did not result in a loss of safety function of the system, the loss of the safety function of a single train for greater than its Technical Specification allowed-outage time, the loss of a safety function of one or more non-Technical Specification trains of equipment designated as risk-significant per 10 CFR 50.65 for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or screen as potentially risk-significant due to seismic, fire, flooding, or severe weather initiating event (Section 4OA2.e(2)iv).

Cornerstone: Occupational Radiation Exposure

Green.

The team documented a self-revealing non-cited violation of Technical Specification 5.7.2.d for the failure to provide an adequate alarming dosimeter to a worker entering a high-high radiation area. Specifically, a radiation worker could not hear an electronic dosimeter alarm in the drywell, which was posted as a high-high radiation area. The alarm was identified by a patrolling health physics technician.

The failure to provide an adequate alarming dosimeter to a worker in a high-high radiation area was a performance deficiency. The finding was more than minor because it impacted the occupational radiation safety cornerstone objective to ensure adequate protection of the workers health and safety from exposure to ionizing radiation. The inspectors determined that the finding was of very low safety significance because it did not involve: (1) As-Low-As-Reasonably-Achievable planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose (Section 4OA2.e(2)v).

B. Licensee Identified Violation A violation of very low safety significance which was identified by Energy Northwest has been reviewed by the inspectors. Corrective actions taken or planned by Energy Northwest have been entered into their corrective action program. This violation and the associated corrective action tracking number are listed in Section 4OA7 of this report.

REPORT DETAILS

4OA2 Identification and Resolution of Problems

The team based the following conclusions, in part, on all issues that were identified in the assessment period, which ranged from October 9, 2003 (the last biennial problem identification and resolution inspection) to the end of the inspection on April 27, 2005.

The issues are divided into two groups. The first group, Current Issues, includes problems that were identified during the assessment period where the performance concern also occurred during the same period. The second group, Historical Issues, includes issues that were identified during the assessment period but the original performance deficiency occurred outside the period of interest.

a.

Effectiveness of Problem Identification

(1) Inspection Scope The inspectors reviewed items selected across the seven cornerstones to determine if problems were being properly identified, characterized, and entered into the corrective action program for evaluation and resolution. The team performed select equipment walkdowns and reviewed operator logs, work orders, plant tracking logs, and action requests for equipment deficiencies that should have also been captured in the corrective action program. The team also reviewed a sample of Energy Northwest audits and self assessments, trending reports, system health reports, and various other reports and documents related to the corrective action program.

The team interviewed station personnel and evaluated corrective action documentation to determine the licensees threshold for identifying problems in their corrective action program. In addition, the team reviewed the licensees evaluation of selected industry experience information, including operator event reports, NRC Generic Bulletins and Information Notices, and generic vendor notifications to assess if issues applicable to Columbia Generating Station were appropriately addressed.

(2) Assessment The licensee generally maintained a healthy problem identification program. In most cases, the licensee properly identified performance trends and appropriately captured applicable industry operating experience in their program. However, the licensee was periodically challenged by their failure to consistently identify their own problems, as the NRC continued to identify noncompliances in this area.

Current Issues Example 1: The NRC identified an inadequate maintenance procedure, in that the procedure instructed plant personnel to take steps which, inadvertently, rendered a safety-related battery and battery charger inoperable. The licensee had failed to identify the improper test configuration during the procedure change process (NRC Inspection Report 05000397/2004002).

Example 2: The NRC identified that operators failed to promptly return an average power range monitor to service following maintenance. Operators missed opportunities to identify the problem during control board walkdowns (NRC Inspection Report 05000397/2004004).

Example 3: The NRC identified that plant personnel failed to promptly identify disengaged 480 VAC breaker seismic latches. 8 out of 21 breakers in a safety-related motor control center did not have their latches fully engaged. The licensee missed prior opportunities to identify the problem during routine breaker inspections (see Section 4OA2.e(2)(i) of this report).

Example 4: The NRC identified that the licensee failed to document several failed electrical disconnects in their corrective action program (NRC Inspection Report 05000397/2004003).

Example 5: The NRC identified that, since breaker installation in 2001, the licensee missed several opportunities to identify seismically nonconforming 4160 VAC breakers.

The breakers did not fit properly in the cubicles and the front wheels lifted off the floor, which was inconsistent with the seismically qualified configuration (see Section 4OA2.e(2)(ii) of this report).

Historical Issues None.

b.

Prioritization and Evaluation of Issues

(1) Inspection Scope The team reviewed condition reports, problem evaluation requests and operability evaluations to assess the licensees ability to evaluate the importance of adverse conditions. The team reviewed a sample of condition reports, apparent cause analyses and root cause analyses to ascertain whether the licensee properly considered the full extent of conditions, generic implications, common causes, and previous occurrences.

In addition, the team reviewed licensee evaluations of selected industry operating experience reports, including licensee event reports, NRC Generic Letters, Bulletins and Information Notices, and generic vendor notifications to assess whether issues applicable to Columbia Generating Station were appropriately addressed.

The team performed a historical review of condition reports and problem evaluation requests written over the last 5 years that addressed Class 1E 4160 V breakers, Barton differential pressure instruments, secondary containment pressure, and normally-energized Struthers-Dunn relays.

(2) Assessment The team concluded that problems were generally prioritized and evaluated in accordance with the licensees corrective action program guidance and NRC requirements. The team found that for the sample of root cause analyses reviewed, the licensee was generally self-critical and thorough in evaluating the causes of significant conditions adverse to quality. Notwithstanding the above, the team observed that, for the past four years, poor evaluations of breaker and switchgear problems still challenge the site, as NRC identified and self-revealing issues continue to surface. Several of the issues involved seismic qualification problems. A similar observation was included in the last problem identification and resolution inspection report.

Current Issues Example 1: The NRC identified that the licensee failed to properly evaluate degraded electrical disconnects, as electrical disconnects continued to unexpectedly pop open due to inadequate maintenance. Further, the licensee also failed to properly address the maintenance issues impact on the disconnects seismic qualification (NRC Inspection Report 05000397/2004003).

Example 2: The NRC identified that the licensee had not properly evaluated the extent of condition for an improperly set Division III diesel fan breaker. The breaker tripped unexpectedly in service. The licensee determined that the failure to perform postmaintenance testing contributed to the problem, but then failed to address four other similar breakers whos settings were adjusted in approximately the same time frame (NRC Inspection Report 05000397/2004003).

Example 3: The NRC identified that the licensee failed to properly evaluate the extent-of-condition for improper 480 VAC breaker over-current trip settings. The licensee had specified corrective actions for breakers associated with one manufacturer but other breakers were also affected. In response to NRC questions, two safety-related breakers and one non-safety related breaker required recalibration (see Section 4OA2.e(2)(iii) of this report).

Historical Issues Example 4: The NRC identified that the licensee failed to take prompt corrective actions to address a condition adverse to fire protection. The protected low pressure coolant injection train was vulnerable to water hammer, since 1997, due to a leaking pump discharge check valve (NRC Inspection Reports 05000397/2003002 and 05000397/2004002).

c.

Effectiveness of Corrective Actions

(1) Inspection Scope The team reviewed plant records, primarily condition reports and problem evaluation requests, to verify that corrective actions related to the issues were identified and implemented, including corrective actions to address common cause or generic concerns. The team sampled specific technical issues to evaluate the adequacy of the licensees operability determinations.

Finally, the team reviewed a sample of condition reports and problem evaluation requests that addressed past NRC identified violations, for each affected cornerstone, to ensure that the corrective actions adequately addressed the issues as described in the inspection reports. The team also reviewed a sample of corrective actions closed to the other condition reports, problem evaluation requests, and other tracking programs to ensure that corrective actions were still appropriate and timely.

In the 2004 end-of-cycle performance assessment, the NRC identified a substantive cross-cutting issue in the area of human performance. The team evaluated the licensees actions to address the substantive cross-cutting issue.

(2) Assessment The licensees implementation of their corrective action program was generally effective.

The NRC identified a few instances where corrective actions were not adequate but, overall, the licensee properly implemented their program.

With respect to the cross-cutting human performance issue, the licensees subsequent self assessments, audits, and third-party assessments were critical and thorough.

While the licensee has observed some improvement in their human performance indicators, the team could not determine the overall adequacy of the corrective measures because the actions were not in place for a sufficient period of time.

Current Issues Example 1: The licensee failed to take effective measures to prevent repetitive failures of a high pressure core spray diesel room fan breaker. The first failure occurred because the specified trip setting was too low. The licensee corrected the trip setting.

However, plant personnel subsequently used the same inappropriate design information at a later date when adjusting the breaker setting. Consequently, the breaker failed in service a second time (self-revealing, NRC Inspection Report 05000397/2004003).

Example 2: The NRC identified that the licensee failed to take adequate corrective measures to address 480 VAC breaker seismic latches following previously identified problems. The licensee did not specify the configuration requirements in plant maintenance documents. During a plant walkdown, the inspectors identified an unlatched (non-conforming) Division III diesel room fan breaker. The disclosure of this issue led to the identification of several other nonconforming breakers (see Section 4OA2.e(2)(iii) of this report).

Historical Issues Example 3:

The licensee identified that their corrective measures to address a previously identified problem with minimum steam cooling reactor water level were inadequate. The licensee had identified that the Emergency Plan, emergency operating procedures, and emergency plan implementing procedures were not properly updated. Corrective measures were inadequate because they only addressed the emergency operating procedures (see Section 4OA7 of this report).

d.

Assessment of Safety-Conscience Work Environment

(1) Inspection Scope The team interviewed more than 20 individuals from the licensees staff, representing a cross section of functional organizations and supervisory and non-supervisory personnel. These interviews assessed whether conditions existed that would challenge the establishment of a safety-conscience work environment. The team also reviewed a plant self-assessment that evaluated, in part, the safety-conscious work environment at Columbia Generating Station.

(2) Assessment The team concluded that a safety-conscience work environment existed at the Columbia Generating Station. The team determined that employees and contractors felt free to enter issues into the corrective action program and raise safety concerns to their supervision, to the employees concern program, and to the NRC. The team received an isolated comment regarding receptiveness of some supervisors to initiating condition reports. The licensee has initiated corrective measures to address the comment. All the interviewees believed that potential safety issues were being addressed.

e.

Specific Issues Identified During This Inspection

(1) Inspection Scope During the reviews described in Sections 4OA2 a.(1), 4OA2 b.(1), 4OA2 c.(1), and 4OA2 d.(1), the team identified the following findings.

(2) Findings and Observations

(i) 480 VAC Breaker Seismic Restraints

Introduction.

The team identified an unresolved item involving the failure to identify 480 V breaker latch and nonconforming seismic restraints (conditions adverse to quality). As a result of the teams inspection, the licensee determined that the Division III diesel generator fuel oil transfer pump breaker was inoperable since it was not in its seismically-qualified configuration.

Description.

In March, 2004 the licensee identified that six breakers in Motor Control Center E-MC-4A were not properly secured to ensure seismic qualification. The breakers should have been secured with a stud, nut and washer assembly, although the assembly was not depicted on design drawings. Two of those breakers controlled the Division III diesel generator room inlet fan and the Division III diesel generator fuel oil transfer pump. The condition was documented in Problem Evaluation Request 204-0604.

While reviewing the corrective actions for the above finding, the team identified a problem which led to the discovery of multiple seismically nonconforming breakers.

During a walkdown of the Division III diesel generator room inlet fan breaker, the team noted that a latch (required for seismic qualification) was not fully engaged. In response to the teams finding, the licensee reviewed the configuration of other breakers in Motor Control Center E-MC-4A and found an additional 8, out of 21 breakers inspected, where at least one of the seismic latches was not properly secured. In addition, the licensee found that the Division III diesel generator fuel oil transfer pump breaker did not have the stud, nut and washer assembly, noted in the first paragraph above. These issues were documented in Condition Reports 2-05-01854 and 02-05-01845. As an additional corrective measure, the licensee planned to inspect the remaining breakers with similar configurations.

The licensee declared most of the breakers degraded but operable. However, the Division III diesel generator fuel oil transfer pump was declared inoperable until a stud assembly could be installed. Plant personnel determined that the stud assembly had been missing since January 5, 2005 when the breaker had been replaced under Work Order 01085125. The work order did not address the installation of the stud assembly following breaker replacement.

These issues had cross-cutting aspects in the areas of problem identification (in that work was in the area of the breakers and workers failed to discover the nonconformances) and effectiveness of corrective actions (in that, following the licensees initial finding concerning the stud nuts, the licensee failed to update procedures to ensure proper installation of the fuel oil transfer pump breakers seismic restraint assembly following maintenance).

Analysis.

The failures to: 1) identify conditions adverse to quality (unlatched seismic restraints); and 2) provide adequate procedures to ensure that seismic restraints were properly restored following maintenance were performance deficiencies. The finding is more than minor because it impacted the mitigating systems cornerstone objective of availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Significance Determination Process Phase 1 worksheet, the team determined that the finding was potentially risk significant because it affected seismic initiating events. The finding requires a Phase 3 analysis, which was not completed at the close of the inspection. This issue is unresolved pending completion of the significance determination.

Enforcement.

The inspectors had not completed the enforcement review at the close of the inspection. This issue is unresolved for compliance pending completion of that review (URI 05000397/2005008-01).

(ii) Safety-Related 4160 VAC Breaker Seismic Qualification

Introduction.

The team documented an unresolved item concerning the seismic qualification of 4160 VAC breakers. The front wheels on several of the safety-related breakers did not touch the floor due to breaker-cubicle fit-up problems, which was not consistent with the seismically tested configuration.

Description.

During the spring, 2001 refueling outage, the licensee replaced 22 existing 4160 VAC Westinghouse DHP-350 breakers with the Westinghouse DHP-VR 350 vacuum-operated breakers manufactured by Cutler-Hammer. Sixteen breakers have a safety function to reposition during design basis accidents, including those postulated accidents involving seismic events. The new breakers were utilized in power circuits for emergency diesel generators, standby service water pumps, and emergency core cooling system pumps. The breakers were installed in the old breaker cubicles. The new breakers did not have exactly the same dimensions as the old breakers, which resulted in cubicle fitup issues. Operations typically had problems getting the breakers properly positioned. The inspectors noted the following related issues:

• On February 13 and 22, 2002, operators initiated Problem Evaluation Requests 202-0476 and 202-0556 to document the excessive manual force to engage the seismic latches for two of the safety-related 4160 VAC Breakers. As part of the resolution, engineers initiated a modification to taper the seismic latches to provide a better fit.

• On December 17, 2003, the licensee initiated Problem Evaluation Request 203-4385 to document that three safety-related 4160 V breakers were bent horizontally across the face approximately 5 inches from the top of the panel.

Excessive force to engage the seismic latches was suspected as the cause of the panel damage.

• On May 17, 2004, an equipment operator noticed that the front wheels for one of the breakers were lifted off the floor, Problem Evaluation Request 204-0775.

This was inconsistent with the seismically tested configuration. The licensee checked the other safety-related 4160 VAC breakers and found that five breakers had both wheels off the floor and eight breakers had one wheel off the floor. The maximum distance between the wheels and the floor was 1/16 of an inch. The licensee initiated Follow-up Assessment of Operability 204-0775 to evaluate operability. The licensee analyzed the as-found configuration and found it acceptable.

The team noted that, normally, electrical equipment required to operate during a seismic event is tested in its qualified configuration, versus qualified by analysis. Engineering personnel stated that if a seismic event were to occur, the breakers would slide down such that all four wheels would touch the floor and the breakers would return to their seismically-qualified configuration. The team requested a technical analysis demonstrating this theory. The team had not completed the NRC evaluation prior to the end of the inspection period. This issue is unresolved pending completion of that evaluation.

The team noted the following historical problems with these breakers:

• NRC Inspection Report 05000397/2002-005, dated June 24, 2002, described a White finding associated with the licensees failure to take effective corrective actions to address design control issues associated with mechanism-operated cell switches for the safety-related 4160 VAC breakers. After, the licensee had experienced numerous operational problems it was determined that they had failed to perform critical maintenance.

• NRC Inspection Report 05000397/2003-009, dated November 24, 2003, described a Green finding associated with the licensees failure to promptly correct a seismic qualification issue associated with safety-related 4160 VAC breaker truck-operated cell position switches. The licensee had experienced eight operational problems before the seismic vulnerability was identified, by the NRC, and corrected.

The team determined that the issue had at least one crosscutting aspect in the area of problem identification. The licensee had missed numerous opportunities to identify the seismically nonconforming breaker configuration prior to 2004, when the condition was eventually found.

Analysis.

A significance determination is not warranted at this time.

Enforcement.

Enforcement will be considered when the NRC completes the evaluation of this issue (URI 05000397/2005008-02).

(iii) Inadequate 480 VAC Breaker Settings

Introduction.

The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XVI (Corrective Actions) for the failure to promptly identify nonconforming breaker settings (conditions adverse to quality). Following a breaker malfunction (tripping too soon), the NRC questioned the adequacy of the licensees extent of condition review. As a result, the licensee identified two additional safety-related breakers and one non-safety related breaker that required calibration. This issue had cross-cutting aspects associated with problem evaluation.

Description.

On August 20, 2003, during a Division III diesel generator surveillance, the diesel room exhaust fan breaker (DEA-42-4A4E) tripped on overcurrent. The licensee determined that plant personnel had failed to properly test the breaker following a previous adjustment. The licensee determined that the fan was not needed as long as outside ambient temperatures were below 105° F. Plant temperatures had not exceeded 105E F since the improper adjustment occurred.

On GE Type TEC motor controller breakers, the overcurrent trip point is set by adjusting a dial on the breaker. The dial has indication of the trip setting. Engineering personnel had assumed that a breakers indicated setting was accurate. However, following the unexpected tripping of Breaker DEA-42-4A4E, engineers learned that the indication was not reliable. As a corrective measure, the licensee verified that the remaining GE Type TEC breakers had acceptable trip values.

Since breakers made by other manufacturers also utilized similar trip dial indicators, the team questioned the adequacy of the licensees extent-of-condition review. More specifically, the team was concerned about Westinghouse DS-416 metal-clad and ITE Gould molded-case circuit breaker settings. In response to the teams questions, the licensee could not verify that the following breakers had acceptable trip settings:

• Safety-related Breaker CCH-CB-CR1A, supply breaker to control room chiller 1A (the chiller itself was non-safety related).

• Safety-related Breaker E-CB-71/7C, feeder breaker from safety Bus E-SL-71 to non-safety bus E-MC-7C.

• Nonsafety-related Breaker E-CB-71/7C, feeder breaker to cooling tower valves.

The licensee subsequently recalibrated the breakers.

Analysis.

The failure to properly evaluate a condition adverse to quality and address the full scope of the problem was a performance deficiency. The finding was more than minor because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Manual Chapter 0609, Appendix A, Significant Determination Process, Phase 1 screening worksheet, this finding had very low safety significance because it did not result in a loss of safety function, a loss of a safety-related train for greater than its Technical Specification allowed outage time, the loss of risk-significant non-Technical Specification trains for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or screen as potentially risk significant due to a seismic, fire, flooding, or severe weather initiating event.

Enforcement.

10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that conditions adverse to quality be promptly identified. Contrary to the above, the licensee failed to promptly identify non-conforming safety-related 480 V breaker settings. Because this issue is of very low safety significance and has been entered into the corrective action program as Condition Report 2-04-01555, this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000397/2005008-03).

(iv) Startup Transformer E-TR-S

Introduction.

The team documented a self-revealing finding to address an inadequate maintenance procedure. The procedure failed to specify the correct relay contact configuration for a startup transformer control circuit relay. When maintenance craftsmen installed the relay and placed the unit in service, the startup transformer tripped off-line. It was recovered approximately 20 minutes later. The issue had crosscutting aspects associated with human performance, adequacy of procedures.

Description.

The startup transformers provide power to safety-related as well as balance of plant systems when the reactor is shutdown. On November 1, 2003, maintenance personnel performed on-line maintenance on Startup Transformer E-TR-S.

The maintenance included calibrating and installing a replacement for Relay E-RLY-50TSN. The relay is part of the overcurrent trip circuitry for the startup transformer. Maintenance personnel did not know that they needed to configure the contacts to the normally-open position prior to installation. In addition, the necessary instructions were not contained in plant procedures or the associate work order.

Consequently, when operators placed the transformer back in service it tripped off-line.

Plant personnel recovered the startup transformer in about 20 minutes. Following the event, the licensee added instructions to Procedure PPM 10.25.21, "Testing & Setting Instantaneous Overcurrent Relays, Work Order WOT 01038854, as well as other procedures, to ensure proper relay configuration in the future.

Analysis.

The failure to provide adequate maintenance instructions for startup transformer work was a performance deficiency. The finding is more that minor because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. In addition, the finding is similar to non-minor Example 5.b in Inspection Manual Chapter 0612, Appendix E, in that the transformer was placed back into service before the problem was corrected. Using the Manual Chapter 0609 Significance Determination Process Phase 1 screening worksheet, the finding was of very low safety significance because the finding did not result in a loss of safety function of the system, the loss of the safety function of a single train for greater than its Technical Specification allowed outage time, the loss of a safety function of one or more non-Technical Specification trains of equipment designated as risk-significant per 10 CFR 50.65 for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or screen as potentially risk-significant due to a seismic, fire, flooding, or severe weather initiating event.

Enforcement.

No violation of regulatory requirements occurred because the affected equipment was nonsafety-related. The licensee documented the problem in Problem Evaluation Requests 203-3940 and 203-4215 (FIN 05000397/2005008-04).

(v) Inadequate Alarming Dosimeter

Introduction.

The team documented a Green self-revealing noncited violation of Technical Specification 5.7.2.d.1 for the failure to provide an adequate alarming dosimeter to a worker entering a high-high radiation area. Specifically, a radiation worker could not hear an electronic dosimeter alarm in the drywell which was posted as a high-high radiation area.

Description.

On June 20, 2003, an individual working in the drywell received an electronic dosimeter dose alarm but could not detect it. The worker was directed to leave the area by a roving radiation protection technician that heard the alarm. This event was documented in Problem Evaluation Request 203-2490. The licensee determined that the worker had entered the drywell on the wrong radiation work permit and had left his hearing aid in the shop. The team noted that the resolution for Problem Evaluation Request 203-2490 was narrow in scope and did not attempt to determine why the co-workers also did not hear the alarm. The licensee initiated Condition Report 2-05-01764 to evaluate this weakness. This issue was self-revealing by the alarming dosimeter.

Analysis.

The failure to provide a worker in a high-high radiation area with an adequate alarming dosimeter was a performance deficiency. The finding was more than minor because it impacted the occupational radiation safety cornerstone objective to ensure adequate protection of the workers health and safety from exposure to radiation from radioactive material. This finding was evaluated with the occupational radiation safety significance determination process in Inspection Manual Chapter 0609, Appendix C.

The inspectors determined that the finding was of very low safety significance because it did not involve: 1) ALARA planning and controls; 2) an overexposure; 3) a substantial potential for overexposure; or 4) an impaired ability to assess dose as defined in Inspection Manual Chapter 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process."

Enforcement.

Technical Specification 5.7.1.d.2 states, in part, that any individual permitted to enter a high radiation area with dose rates greater than 1.0 rem/hour (high-high radiation area) shall be provided with a radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Contrary to the above, on June 20, 2003, the licensee failed to provide an adequate alarming dosimeter to a worker entering a high-high radiation area. The dosimeter was inadequate because the worker could not hear the alarm. Because this violation was of very low safety significance and was entered into the corrective action program as Problem Evaluation Request 203-2490, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy (NCV 05000397/2005008-05).

4OA6 Exit Meetings

On March 31, 2005, the team discussed the preliminary findings with Mr. J. V. Parrish, Chief Executive Officer and other members of the licensees staff. On April 27, 2005, the team leader conducted the final exit meeting with Mr. Roberto Torres, Quality Assurance and Corrective Action Manager, and other members of the licensees staff by telephone. The team asked the licensee whether any materials examined during the inspection should be considered proprietary. Proprietary information was reviewed by the team but all the proprietary information was returned to the licensee prior to the close of the inspection. No proprietary information was disclosed in this report.

4OA7 Licensee Identified Violations

The following violation of very low risk significance (Green) was identified by the licensee. The violation meets the criteria of Section VI of the NRC Enforcement Policy for being dispositioned as a noncited violation.

• 10 CFR 50.54(q) requires, in part, that a licensee shall follow and maintain in effect emergency plans which meet, in part, the requirements in 10 CFR 50, Appendix E. This regulation requires that provision be employed to ensure that the emergency plan and its implementing procedures are maintained up to date.

Contrary to the above, following a core design change, the Emergency Plan and the emergency plan implementing procedures (as well and the emergency operating procedures) were not maintained up to date. The documents should have been changed before the reactor startup following the design change. The design change affected the minimum steam cooling reactor water level, which is lowest acceptable water level that still ensures adequate core cooling. This value is utilized as a juncture for determining protective measures for the public and emergency actions in the control room. At a later date, the licensee also identified that initial corrective measures to address the issue were inadequate in that some of the documents were still not properly updated. The licensee subsequently performed an analysis and determined that the previously specified minimum steam cooling reactor water level ensured adequate core cooling for the affected periods of time. The revised value was not needed until the end of core life, which had not yet occurred. Therefore, the issue was of very low safety significance because the fuel clad remained operable consistent with Generic Letter 91-18, Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions, Revision 1.

This issue had crosscutting aspects in the area of effectiveness of corrective actions.

ATTACHMENT PARTIAL LIST OF PERSONS CONTACTED Licensee V. Parrish, Chief Executive Officer/Chief Nuclear Officer D. Atchinson, Vice President, Nuclear Generation S. Belcher, Manager, Operations J. Bekhazi, Manager, Maintenance I. Borland, Manager, Radiation Protection D. Coleman, Manager, Regulatory Programs T. Lynch, Plant General Manager S. Oxenford, Vice President, Technical Services R. Torres, Manager, Quality Assurance and Corrective Action C. Whitcomb, Vice President, Organizational Performance and Staffing NRC R. Cohen, Resident Inspector, Columbia Generating Station G. Replogle, Senior Reactor Inspector, Engineering Branch 2 LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened and Closed 05000397/2005008-03 NCV Failure to Evaluate the Extent of Condition for 480 V Breaker Overcurrent Knob Settings (Section 4OA2.e(2)(iii))05000397/2005008-04 FIN Failure to Maintain Adequate Maintenance Procedure Resulted in Startup Transformer Tripping (Section 4OA2.e(2)(iv))05000397/2005008-05 NCV Failure to Provide an Adequate Alarming Dosimeter (Section 4OA2.e(2)(v))

Opened 05000397/2005008-01 URI Failure to Identify and Correct 480 V Breaker Seismic Restraint Issues (Section 4OA2.e(2)(i))05000397/2005008-02 URI Failure to Identify and Correct a Seismically Nonconforming Configuration Related to Safety-Related 4160 V Breakers (Section 4OA2.e(2)(ii))

LIST OF ACRONYMS ADAMS Agency Document And Management System CFR Code of Federal Regulations FIN finding NCV noncited violation NRC Nuclear Regulatory Commission PARS Publicly Available Records System URI unresolved item

DOCUMENTS REVIEWED

Action Requests

3940 5806 6321 6961 6964 7142

Condition Reports

2-04-00029 2-04-02506 2-04-03362 2-04-04716 2-04-05425

2-04-00149 2-04-02577 2-04-03450 2-04-04881 2-04-05538

2-04-00152 2-04-02783 2-04-03558 2-04-05160 2-04-05594

2-04-00350 2-04-02973 2-04-03577 2-04-05165 2-04-05597

2-04-00353 2-04-03172 2-04-03609 2-04-05177 2-04-05617

2-04-00464 2-04-03176 2-04-03621 2-04-05267 2-04-05633

2-04-00893 2-04-03351 2-04-03647 2-04-05321 2-04-05698

2-04-01023 2-04-03701 2-04-05922

2-04-01035 2-04-03703 2-04-06058

2-04-01107 2-04-03987 2-04-06072

2-04-01130 2-04-04038 2-04-06224

2-04-01224 2-04-04178 2-04-06294

2-04-01292 2-04-04209 2-04-06340

2-04-01360 2-04-04223 2-04-06363

2-04-01453 2-04-04248 2-04-06423

2-04-01534 2-04-04264 2-04-06472

2-04-01833 2-04-04333 2-04-06491

2-04-02290 2-04-04395 2-04-06493

2-04-02314 2-04-04565 2-04-06501

2-04-02458 2-04-04622 2-04-06697

Attachment

2-04-06720 2-05-00788 2-05-01719

2-04-06745 2-05-00803 2-05-01720

2-04-06853 2-05-00920 2-05-01721

2-05-00014 2-05-00945 2-05-01722

2-05-00108 2-05-01072 2-05-01723

2-05-00233 2-05-01231 2-05-01724

2-05-00482 2-05-01341 2-05-01725

2-05-01404 2-05-01726

2-05-01458 2-05-01727

2-05-01468 2-05-01728

2-05-01530 2-05-01729

2-05-01555 2-05-01730

2-05-01559 2-05-01731

2-05-01618 2-05-01732

2-05-01679 2-05-01733

2-05-01690 2-05-01734

2-05-01691 2-05-01751

2-05-01695 2-05-01799

2-05-01709 2-05-01801

2-05-01711 2-05-01802

2-05-01712 2-05-01803

2-05-01713 2-05-01808

2-05-01714 2-05-01809

2-05-01715 2-05-01818

2-05-01716 2-05-01845

2-05-01717 2-05-01854

2-05-01718

Attachment

Calculation Modification Requests

2902 2716 588

Drawings

E502-2, Main One Line Diagram, Revision 52

E503-7, Auxiliary One Line Diagram, Revision 76

FSKC-XX-0001-0-096, Seismic Locking Bar for Vacuum Circuit Breaker, Revision 0

M521-1, "Flow Diagram Residual Heat Removal System Loop A," Revision 98

M521-2, "Flow Diagram Residual Heat Removal System Loop B," Revision 99

Plant Tracking Logs

A 139934 A 209873 A 216444 A 218433 A 220696

A 206427 A 210048 A 216797 A 218437 A 221093

A 207835 A 211471 A 216869 A 218441 A 221094

A 207871 A 211473 A 217061 A 219495 A 221095

A 208074 A 211941 A 217937 A 218508 A 221096

A 208289 A 212289 A 217942 A 218509 A 221180

A 208370 A 213384 A 218129 A 218524 A 221182

A 208566 A 213779 A 218423 A 218525 A 223269

A 208780 A 214002 A 218424 A 218526 A 223611

A 208986 A 214503 A 218426 A 219802 A 224403

A 209189 A 214947 A 218427 A 219803 A 224418

A 209435 A 215422 A 218431 A 220547 A 224495

A 209799 A 215493

Problem Evaluation Reports

294-0762 202-2575 203-0956 203-2225 203-3145

297-1003 202-2707 203-0978 203-2254 203-3174

299-0127 202-2984 203-1003 203-2295 203-3181

200-0025 202-3359 203-1094 203-2326 203-3182

201-1014 202-3463 203-1263 203-2382 203-3183

201-2897 202-3581 203-1271 203-2384 203-3184

2-0556 203-0288 203-1560 203-2396 203-3185

2-0476 203-0332 203-1561 203-2534 203-3187

2-1365 203-0361 203-1580 203-2565 203-3188

2-1408 203-0423 203-1683 203-2732 203-3196

2-1522 203-0427 203-1749 203-2901 203-3204

2-1687 203-0428 203-1783 203-2955 203-3262

2-1800 203-0465 203-1902 203-2966 203-3263

2-2026 203-0500 203-2123 203-3036 203-3283

2-2199 203-0510 203-2127 203-3050 203-3287

2-2235 203-0607 203-2132 203-3059 203-3288

2-2306 203-0614 203-2202 203-3104 203-3289

2-2558 203-0725 203-2208 203-3129 203-3290

Attachment

Problem Evaluation Reports (Cont.)

203-3365 203-3955 204-0334 204-0780 204-1015

203-3366 203-3975 204-0353 204-0783 204-1020

203-3370 203-3996 204-0376 204-0789 204-1024

203-3373 203-4001 204-0386 204-0795 204-1041

203-3388 203-4003 204-0457 204-0804 204-1042

203-3411 203-4041 204-0469 204-0811 204-1047

203-3416 203-4049 204-0575 204-0822 204-1056

203-3423 203-4064 204-0594 204-0825 204-1079

203-3434 203-4210 204-0595 204-0842 204-1093

203-3459 203-4211 204-0604 204-0848 204-1114

203-3460 203-4228 204-0608 204-0857 204-1127

203-3471 203-4231 204-0610 204-0858 204-1129

203-3491 203-4244 204-0612 204-0870 204-1165

203-3513 203-4330 204-0628 204-0881 204-1166

203-3533 203-4385 204-0644 204-0883 204-1172

203-3535 203-4387 204-0646 204-0906 204-1177

203-3552 203-4482 204-0647 204-0929 204-1186

203-3559 203-4493 204-0660 204-0935 204-1187

203-3561 204-0012 204-0673 204-0951 204-1211

203-3578 204-0015 204-0701 204-0954 204-1219

203-3606 204-0108 204-0711 204-0959 204-1225

203-3617 204-0112 204-0718 204-0961 204-1227

203-3645 204-0116 204-0719 204-0973 204-1237

203-3684 204-0118 204-0737 204-0981 204-1266

203-3689 204-0133 204-0745 204-0982 205-0024

203-3693 204-0189 204-0746 204-0983 205-0057

203-3766 204-0197 204-0766 204-0984 205-0064

203-3852 204-0287 204-0768 204-0986 205-0136

203-3872 204-0302 204-0775 204-0987 205-0199

203-3940

Plant Procedures

Procedure Title Revision

CCM-1-1 Vibration Analysis and Trending 0

OSP-CAC-B701 CAC-HR-1A Preheater Operability Test 7

OSP-CAC-B702 CAC-HR-1B Preheater Operability Test 7

OSP-ELEC-M701 Diesel Generator 1 - Monthly Operability Test 19

OSP-ELEC-W102 Electrical Distribution Subsystem Breaker Alignment and 12

Power Availability Verification

OSP-HPCS-A701 High Pressure Core Spray Keep Fill Integrity Test 0

Attachment

OSP-HPCS/IST-Q701 HPCS System Operability Test 23

OSP-MS/IST-Q701 MSIV Closure Test - Shutdown 7

PPM 2.10.1 System Operating Procedures, Heating, Ventilation, and 28, 32

Air Conditioning

PPM 10.1.13 Foreign Material Controls for Systems and Components 19

PPM 10.17.3 Main Steam Isolation Valve Overhaul 16

PPM 10.25.20 Testing & Setting Electromechanical Time Overcurrent 15

Relays

PPM 10.25.35 Testing and Setting GE CEH Relays 10 - 11

PPM 11.2.10.17 Operation of the NMC Continuous Air Monitor 0

PPM 11.2.13.8 Airborne Radioactivity Surveys 6

SPIP-SEC-08 Issue, Receipt, Storage, Maintenance, Accountability 6

and Handling of Security Equipment, Firearms and

Ammunition

SPIP-SEC-20 Central Alarm Station Access Instructions 10

SWP-CAP-01 Corrective Action Program 9

SWP-CAP-02 Cause Determination 2

SWP-CAP-03 Operating Experience Program 3

SWP-CAP-05 Corrective Action Review Board (CARB) 2

SWP-CAP-06 Condition Review Group (CRG) 2

SWP-CAP-07 Department Assessment Review Program 2

SWP-SEC-02 Protection of SAFEGUARDS Information 5

SWP-SEC-03 Vital & Protected Area Personnel/Vehicle Access 13

Controls and Security Responsibilities of Site Personnel

Other

Calculation ME-02-90-16, Pressure Drop Verification for LPCS System, Revision 1

Calculation NE-02-39, Evaluation of CAC Performance Under LOCA Conditions,

April 5, 2004

Design Specification for Division 300 Section 309 Standby Service Water System, Revision 7

Equipment Qualification Group Task W01425, Breaker Operability, May 18, 2004

Firearms and Range Safety Course, Squad Attendance Rosters, March 15 thru

April 29, 2004

Attachment

GE SIL No. 477, Main Steam Isolation Valve Closure, December 13, 1988

Licensing Document Change Notice FSAR-03-076

Memorandum SS2-PE-95-469, Examination of Agastat and Struthers-Dunn Relays

Removed During R10, August 30, 1995

Memorandum, Closure of EOR 81043GA-3/4, Evaluation of GE SIL #477, Main Steam

Isolation Valve Closure, July 18, 1990

NE-02-39, Evaluation of CAC Performance Under LOCA Conditions, April 5, 2004

Nuclear Energy Institute 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1

NRC Information Notice 80-16, Shaft Seal Packing Causes Binding in Main Steam Swing

Disc Check and Isolation Valves

NRC Information Notice 88-51, Failure of Main Steam Isolation Valves

NRC Information Notice 03-17, Reduced Service Life of Automatic Switch Company (ASCO)

Solenoid Valves with Buna-N Material

NRC Information Notice 03-18, General Electric Type SBM Control Switches with Defective Cam

Followers

Nuclear Security Officer's Course, Range Operations Lesson Guide, Revision 2

Quality Assurance Audit AU-CA-03, Corrective Action Program Audit

Security Training Observation Checklist, Firing Range, No date

Security Training Range Operating Instructions, date January 12, 2005

Security Training Range Operating Instructions, date October 14, 2003

Self Assessment SA-2003-0046, Utilities Service Alliance Safety Culture Assessment

Self Assessment SA-2004-0056, Operations Self-Evaluation

Self Assessment SA-2004-0060, Nuclear Safety Culture Effectiveness Review

Work Order Task 00SKD3, E-CB-71/7C Vendor Remove/Replace

Work Order Task 00SKD6, CCH-CB-CR1A Vendor Remove/Replace

Attachment

Work Orders

00SF08 01041280 01041914 01069867 01077291

00SKD3 01041631 01044697 01075433 01081111

00SKD6 01041744 01064035 01075434 01085125

01038854 01041763 01068300 01075837 01085177

01040790 01041896

Work Requests

29018809 29032714 29038018 29044289 29045786

29026658 29036287 29038796 29044952

Attachment