IR 05000482/2004006
| ML043150521 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/10/2004 |
| From: | Laura Smith Division of Reactor Safety IV |
| To: | Muench R Wolf Creek |
| References | |
| IR-04-006 | |
| Download: ML043150521 (25) | |
Text
ber 10, 2004
SUBJECT:
WOLF CREEK GENERATING STATION -- NRC PROBLEM IDENTIFICATION AND RESOLUTION INSPECTION REPORT 05000482/2004006
Dear Mr. Muench:
On September 29, 2004, the NRC completed an inspection of problem identification and resolution at your Wolf Creek Generating Station. The enclosed report documents the inspection findings, which were discussed with you and other members of your staff on August 6, 2004, and on September 29, 2004.
This inspection was an examination of 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 200 Performance Improvement Requests (PIRs) program documents, apparent and root cause analyses and plant procedures for the identification and resolution of problems. In addition, the team reviewed cross-cutting aspects of NRC and licensee-identified findings and interviewed personnel regarding the safety conscious work environment.
On the basis of the sample selected for review, the team concluded that in general, problems were properly identified, evaluated and corrected. Your processes to identify, prioritize, evaluate, and correct problems were generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality. The team concluded that a positive safety-conscious work environment exists at Wolf Creek.
This report documents two findings that were evaluated under the risk significance determination process as having very low safety significance (Green). The NRC has also determined that violations were associated with these findings. The violations are being treated as noncited 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 significance of these noncited violations, 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, 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 Wolf Creek Generating Station facility.
Wolf Creek Nuclear Operating Corporation -2-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response will be made 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//
Linda Joy Smith, Chief Plant Engineering Branch Division of Reactor Safety Docket: 50-482 License: NPF-42 Enclosure:
Site Vice President Wolf Creek Nuclear Operating Corp.
P.O. Box 411 Burlington, KS 66839 Jay Silberg, Esq.
Shaw, Pittman, Potts & Trowbridge 2300 N Street, NW Washington, DC 20037 Supervisor Licensing Wolf Creek Nuclear Operating Corp.
P.O. Box 411 Burlington, KS 66839 Chief Engineer Utilities Division Kansas Corporation Commission 1500 SW Arrowhead Rd.
Topeka, KS 66604-4027 Office of the Governor State of Kansas Topeka, KS 66612 Attorney General 120 S.W. 10th Avenue, 2nd Floor Topeka, KS 66612-1597
Wolf Creek Nuclear Operating Corporation -3-County Clerk Coffey County Courthouse 110 South 6th Street Burlington, KS 66839-1798 Chief, Radiation and Asbestos Control Section Kansas Department of Health and Environment Bureau of Air and Radiation 1000 SW Jackson, Suite 310 Topeka, KS 66612-1366 Frank Moussa, Technological Hazards Administrator Department of the Adjutant General 2800 SW Topeka Blvd.
Topeka, KS 66611-1287
Wolf Creek Nuclear Operating Corporation -4-Electronic distribution by RIV:
Regional Administrator (BSM1)
DRP Director (ATH)
DRS Director (DDC)
Senior Resident Inspector (FLB2)
Resident Inspector (TBR2)
SRI, Callaway (MSP)
Branch Chief, DRP/B (DNG)
Senior Project Engineer, DRP/B (RAK1)
Staff Chief, DRP/TSS (KMK)
RITS Coordinator (KEG)
Matt Mitchell, OEDO RIV Coordinator (MAM4)
WC Site Secretary (SLA2)
ADAMS: / Yes G No Initials: ______
/ Publicly Available G Non-Publicly Available G Sensitive / Non-Sensitive DOCUMENT: R\_WC\WC2004-06RP-RAK.wpd RIV:DRS\C:PEB DRP/B:SPE DRP/A:SRI DRP/E:PE DRP/B:RI LJSmith RAKopriva DBAllen DStearn TRhoades
/RA/ /RA/ /RA/ via T /RA/ via E /RA/ via E 11/4/04 10/29/04 11/ /04 11/8 /04 11/5/04 C:DRP/B C:PEB DNGraves LJSmith
/RA/ /RA/
11/10/04 11/10 /04 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-482 License: NPF- 42 Report: 05000482/2004006 Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane NE Burlington, Kansas Dates: July 19 - September 29, 2004 Inspectors: Ronald A. Kopriva, Senior Project Engineer Donald B. Allen, Senior Resident Inspector, Comanche Peak Donald L. Stearns, Project Engineer Travis B. Rhoades, Resident Inspector, Wolf Creek Approved by: L. J. Smith, Chief Plant Engineering Branch Division of Reactor Safety Enclosure
SUMMARY
OF ISSUES
IR 05000482/2004-006; 07/19 - 09/29, 2004; Wolf Creek Generating Station; biennial baseline inspection of the identification and resolution of problems. Violations were identified related to simulator fidelity and design control.
The inspection was conducted by a senior project engineer, a senior resident inspector, a resident inspector, and a project engineer. Two green findings of very low safety significance were identified during the inspection and were classified as noncited violations. The significance of most findings is indicated by their color (green, white, yellow, red) using IMC 0609, Significance Determination Process. Findings for which the significant determination process does not apply may be green or 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 200 Performance Improvement Requests program documents, apparent and root cause analyses and plant procedures for the identification and resolution of problems. Based on this review, the team found that the processes to identify, prioritize, evaluate, and correct problems were generally effective; thresholds for identifying issues remained appropriately low and, in most cases, corrective actions were adequate to address conditions adverse to quality.
Cross-cutting aspects, associated with identification, prioritization and evaluation and correction of degraded conditions in the plant were identified. The team found that these cross-cutting aspects were the exception and not the rule and most issues were minor. However, in a few cases, licensee personnel did not initiate corrective action documents for known equipment degradations. In other cases, planned corrective actions were not managed to a satisfactory completion. Either the issue was not corrected by the planned actions, or the planned actions were cancelled.
Based on the interviews, the team concluded that a positive safety-conscious work environment exists at Wolf Creek. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program.
A. Inspector-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green: A self-revealing, noncited violation of CFR 55.46 (1) was identified regarding simulator response to a transient condition. While completing immediate actions following a reactor trip that occurred on February 13, 2004, the Balance of Plant Operator (BOP) observed what he understood to be a malfunction of the steam dump valves. Subsequent investigation revealed that the plant systems operated properly but that the Balance of Plant Operator did not expect the Steam Generator Atmospheric
Relief Valves (ARV) to be open while the steam dumps were closed shortly following a plant trip. The licensee identified that the simulator had not accurately modeled steam generator atmospheric relief valves post-trip operation since initial licensing.
Based on the results of a Significance Determination Process (SDP) using Manual Chapter (MC) 0609, Appendix I, this finding was determined to have very low safety significance, since it involved a simulator fidelity issue which impacted operator actions.
The failure to adequately model plant response in the simulator, discovered on February 19, 2004, is a violation of 10 CFR 55.46(c). This violation is being treated as a noncited violation 05000482/2004006-01 consistent with Section VI.A of the NRC Enforcement Policy (Section 4OA2e).
- Green.
A self-revealing noncited violation of 10 CFR 50, Appendix B, Criterion III, for the failure to assure that design criteria had adequately been translated into specifications and procedures associated with the Emergency Diesel Generators.
Specifically, in December 2002, and February 2003, the licensee failed to correctly adjust the overcurrent trip setpoints on the newly installed, different manufacture,
Emergency Diesel Generator supply fan breakers. On March 12, 2003, Emergency Diesel Generator "A" supply fan Breaker NG03DBF6 was found tripped, but no problem was identified. On April 12 and April 15, 2003, additional failures of NG03DBF6 were identified. Evaluation determined that new breakers had been installed with overcurrent trips set too low to allow for the starting inrush current. The Emergency Diesel Generators were determined not to be affected because the outside temperature had not exceeded 79 degrees Fahrenheit (F), which is the temperature at which the fans are required to be operable.
The finding is greater than minor because it affected that Mitigating Systems Cornerstone objective of equipment reliability, in that the failure of the Emergency Diesel Generator supply fans could have made the Emergency Diesel Generator inoperable if the outside temperatures had exceeded 79 degrees F. The finding is of very low safety significance because at the time of the breaker failures the outside air temperature had not exceeded 79 degrees F; therefore there was no loss of safety function. This violation is being treated as a noncited violation 05000482/2004006-02 consistent with Section VI.A of the NRC Enforcement Policy (Section 4OA2.e).
Licensee-Identified Violations
Violations of very low significance which were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. These violations and corrective actions are listed in Sections 4OA7.
REPORT DETAILS
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
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. Specifically, the teams review included a selection of approximately 200 Performance Improvement Requests. The majority were opened or closed since the last NRC Problem Identification and Resolution Inspection completed on May 17, 2002. The team also performed a historical review of Performance Improvement Requests written over the last five years for the essential service water system, component cooling water system, radiological controls, main power transformers, and the emergency diesel generators. The team reviewed a sample of licensee audits and self assessments, trending reports, system health reports, and various other reports and documents related to the problem identification and resolution program. The audit and self-assessment results were compared with the self-revealing and NRC-identified issues to determine the effectiveness of the audits and self assessments.
The team interviewed station personnel and evaluated corrective action documentation to determine the licensees threshold for identifying problems and entering them into the corrective action program. The team attended morning meetings to evaluate the licensees evaluation of plant issues against corrective action program criteria for Performance Improvement Request initiation. The team evaluated the licensees efforts in establishing the scope of problems by reviewing control room operator logs, security and radiation protection logs and maintenance items.
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 Wolf Creek were appropriately addressed.
A listing of specific documents reviewed during the inspection is included in the attachment to this report.
- (2) Assessment The team determined that, in general, problems were adequately identified and entered into the Performance Improvement Request program. The threshold for entering issues into the Performance Improvement Request program was appropriately low. Recent conditions adverse to quality identified in various logs or other programs were properly entered into the licensees Performance Improvement Request program. However, the team noted several cross-cutting aspects related to problem identification: The inspectors observed that longstanding adverse conditions related to simulator fidelity and tone alert radio distribution had only recently been identified by the licensee. In addition, the inspectors recently identified longstanding fire barrier seal degradations; and there were two examples where the licensee did not initiate a corrective action document for known equipment degradations.
Example 1 - Failure to Promptly Identify a Simulator Fidelity Concern The team determined problem identification related to this self-revealing issue was not timely because of the length of time the issue existed, prior to identification. From initial plant operation until February 19, 2004, the licensee failed to identify that the simulator response to a normal reactor trip differed from actual plant response (Section 4OA2e).
Example 2 - Failure to Promptly Identify Concern with Distribution of Tone Alert Radios per Emergency Plan Design Commitments to FEMA The team determined problem identification related to this licensee-identified issue was not timely based on a number of opportunities that the licensee had to identify a concern with the distribution of tone alert radios for emergency preparedness. Previous operating experience from Callaway and Arkansas indicated a concern with proper controls associated with the distribution of tone alert radios (Section 4OA7).
Example 3 - The Failure to Promptly Identify Long-standing Degraded 3-hour Fire Rated Fire Barrier Seals The team reviewed the circumstances around NRC Inspection Report Noncited Violation 05000482/2004002-02 and found that the licensee had failed to identify problems with fifteen fire barrier seals for a number of years.
Example 4 - Failure to Enter a Degraded Condition Related to Feedwater Regulating Valves into the Corrective Action Program The team reviewed the circumstances around NRC Inspection Report Finding 05000482/2004002-01 and found that maintenance and engineering personnel were aware of a degraded condition and failed to enter it into the corrective action program.
Example 5 - Failure to Enter a Degraded Condition Related to Emergency Diesel Generator Heat Exchanger Tubes into the Corrective Action Program The team reviewed the circumstance around NRC Inspection Report Noncited Violation 05000482/2002004-01 and found that the licensee failed to enter known equipment degradations into their corrective action program. The eddy current testing technician identified five intercooler tubes with indications of less than 30 percent remaining wall and three tubes with absolute drift indications. This condition existed from December 13, 2001 to January 4, 2002, without being entered into the corrective action program.
b. Prioritization and Evaluation of Issues
- (1) Inspection Scope The team reviewed Performance Improvement Requests, engineering operability evaluations and operations operability determinations to assess the licensees ability to evaluate the importance of the conditions adverse to quality. The team reviewed the results of Performance Improvement Request review group meetings that assigned significance and priority to the Performance Improvement Requests. The team reviewed a sample of failure mode analyses, apparent cause analyses and root cause analyses, to ascertain whether the licensee identified and considered the full extent of conditions, generic implications, common causes, and previous occurrences. The team also observed management oversight of the significant conditions adverse to quality including one Corrective Action Review Board meeting.
In addition, the inspectors reviewed licensee evaluations of selected industry operating experience information, including operating event reports and NRC and generic vendor notices, to assess whether issues applicable to Wolf Creek Generating Station were appropriately addressed. The team performed a historical review of Performance Improvement Request reports covering the last five years regarding the high pressure safety injection system, the emergency feedwater system, safety-related battery chargers and the emergency diesel generators to determine if the licensee had appropriately addressed long-standing issues and those that might be age dependent.
A listing of specific documents reviewed during the inspection is included in the attachment to this report.
- (2) Assessment The team concluded that problems were generally prioritized and evaluated in accordance with the licensees Performance Improvement Request program guidance and NRC requirements. The team found that for the sample of root cause analyses reviewed, that the licensee was generally self critical and exhaustive in its research into the history of significant conditions adverse to quality However, the team noted some unrelated cross-cutting aspects related to problem evaluation and prioritization. Examples included the inadequate use of operating experience related to radiation postings, failure to establish an eddy current testing acceptance criteria, a failure to promptly find the cause of unexpected Emergency Diesel Generator supply fan breaker trips, and the failure to fully document equipment degradation which resulted in ineffective troubleshooting.
Example 1 - Failure to Post Correct Radiation Levels During review of the circumstances surrounding noncited violation 05000482/3003006-03, the inspector noted that dose rates had increased in the normal charging pump room and the volume control tank valve galley of the auxiliary building. The health physics staff responded and identified that both rooms had general radiation levels greater than 100 millirem per hour, requiring the areas to be posted as high radiation areas. During the investigation of the event, the licensee also identified that the seal water heat exchanger room radiation levels had increased, which required the area to be posted as a high radiation area. The cause for the elevated dose rates were from the chemical flush of the reactor coolant system. Once identified, the licensee took appropriate timely actions to properly control the areas.
There is sufficient OE and site experience to identify and evaluate the appropriate times when changing dose rates may take place. Operations and Radiation Protection department personnel concurred that, had they discussed the evolution, they would have been able to identify areas of concern and provide the appropriate postings.
Example 2 - Failure to Properly Evaluate Allowable Emergency Diesel Generator heat Exchanger Tube Wall Thinning The team reviewed the circumstance around NRC Inspection Report Noncited Violation 05000482/2002004-01 and found that the licensee failed to provide definitive acceptance criterion for the eddy current testing on the Emergency Diesel Generator A heat exchangers. The failure to have definitive acceptance criterion in the work order as part of the planning process led to a significant delay in evaluating and recognizing the degraded condition of the Emergency Diesel Generator.
Example 3 - Failure to Properly Evaluate a Degraded Light Socket Contributed to a Small Fire When the site watch noticed that the power available indication on the main transformer, phase C power supply/control cabinet was extinguished, he initiated a work request, but did not adequately describe the evidence of shorting on the bulb in the work request. As a result, the on-coming maintenance personnel did not adequately evaluate and correct the problem. Specifically, the Work Request initiator failed to preserve the physical evidence (light bulb which appeared to be shorted in its socket) and failed to produce an adequate written description of the condition as required by the work request procedure.
This minor violation contributed to a small fire in the Main Transformer C Phase control panel.
Example 4 - Inadequate Cause Determination for an Emergency Diesel Generator Supply Fan Breaker Trip Resulted in Repeat failures A self-revealing violation of 10 CFR 50, Appendix B, Criterion III was identified for failure to assure that design criteria had adequately been translated into specifications and procedures associated with the Emergency Diesel Generator supply fan breakers. In December 2002 and February 2003, the Emergency Diesel Generator supply fan breakers were replaced due to parts obsolescence. On 03/12/03, the Emergency Diesel Generator supply fan breaker was found tripped, but an adequate cause evaluation was not performed and no problem was identified. On 04/12/03 and 04/15/03, additional failures were identified. Evaluation determined that new breakers had been installed with overcurrent trips set too low to allow for the starting inrush current. The Emergency Diesel Generators were determined not to be affected because the outside air temperature had not exceeded 79 degrees F, which is the temperature at which the fans are required to be operable (Section 4OA2e).
c. Effectiveness of Corrective Actions
- (1) Inspection Scope The team reviewed approximately 200 condition reports to verify that corrective actions related to the issues were identified and implemented in a timely manner commensurate with safety, including corrective actions to address common cause or generic concerns.
The team reviewed corrective actions planned and implemented by the licensee and sampled specific technical issues to determine whether adequate decisions related to structure, system, and component operability were made.
In addition, the team reviewed a sample of those Performance Improvement Request reports written to address NRC inspection findings to ensure that the corrective actions adequately address the issues as described in the inspection report writeups. The team also reviewed a sample of corrective actions closed to other Performance Improvement Request reports and programs, such as work requests, to ensure that the condition described was adequately addressed and corrected.
A listing of specific documents reviewed during the inspection is included in the attachment to this report.
- (2) Assessment The processes to correct problems were generally effective; in most cases, corrective actions were adequate to address conditions adverse to quality. However, planned corrective actions were not always managed to a satisfactory completion. Six examples of cross-cutting aspects were identified associated with correction of degraded conditions in the plant. The team found corrective actions developed for self-assessments, operating experience and NRC noncited violations, which were not managed to satisfactory completion. Either the issue was not corrected by the planned actions, or the planned actions were cancelled.
Example 1 - Failure to Manage Corrective Action for 1995 Self Assessment Performance Improvement Request 2004-0132 documented that a preventative maintenance activity to replace the Emergency Diesel Generator Main Air Start Distributor air filters was not created in 1996, because the Performance Improvement Request 95-2413 item # 5 was closed to Performance Improvement Request 96-0682, and that Performance Improvement Request failed to capture the action. As a result, a recommended PM from a 1995 self assessment to address the health risk assessment of the Emergency Diesel Generator to replace the air filter, had not been implemented since Refueling Outage RF8. In the mean time the filter has not been changed in 7 1/2 -
8 years. This licensee-identified issue is discussed in Section 4OA7.
Example 2 - Inadequate Corrective Action For Maintaining Residual Heat Removal Systems Operable During Refueling While the Vessel Internals Remain Installed.
The inspectors identified poor implementation of operating experience. While in Mode 6 with water level greater than or equal to 23 feet above the vessel flange, only one train of RHR is required by Technical Specification 3.9.5. However, even with 23 feet of water over the vessel flange, with the upper internals installed there may be insufficient transfer of heat to prevent boiling in the core if forced flow is lost. During this plant condition, both trains of RHR should remain available. To address this issue, the procedure was revised to incorporate a caution, but not in an appropriate location to ensure that the operater would recognize that both trains of RHR would be required, even with greater than or equal to 23 feet of water. This was a minor violation.
Example 3 - Ineffective Corrective Action Management for Failure to Normalize Reactor Vessel Level Indicating System (RVLIS), that Resulted in Both Trains being Inoperable for an Extended Period Licensee Event Report 2003-002, and PIR 2003-0805 documented that both trains of reactor vessel level indicating system (RVLIS) had been inoperable for longer than allowed by Technical Specifications 3.3.3. Based on a review of the licensees planned actions, the LER was closed in NRC Inspection Report 2003-006 as a minor violation.
The Root Cause Analysis Report for PIR 2003-0805 documented that PIR 97-1983 was written to address NRC Information Notice 97-25 dealing with a Diablo Canyon experience with RVLIS. In response to that Performance Improvement Request, Westinghouse reviewed data from RF9 and recommended the system be normalized due to hydraulic changes in the reactor coolant system. PIR 97-1983 was closed with work package 126425 to track the issue, but the work package was closed on July 23, 1999 with no field work complete.
Example 4 - Failure to Properly Evaluate a Significant Condition Adverse to Quality when Revising an Emergency Operating Procedure Due to RVLIS Failed High.
The team reviewed the circumstance around NRC Inspection Report Noncited Violation 05000482/2003004-01 and found that the licensee failed to appropriately evaluate the impact of the reactor vessel level indicating system (RVLIS) being inoperable when questioned by the resident inspector. They did not initially realize that Procedure EMG C-11, Loss of Emergency Coolant Recirculation, Revision 14, could not have been satisfactorily completed with RVLIS inoperable. The procedure reviewer's flawed evaluation was based on incorrect procedure usage rules.
Example 5 - Inadequate Corrective Actions for NRC-identified Violation 05000482/2002008-01 Related to Control Room Evacuation Critical Timeline The team reviewed the circumstance around NRC Inspection Report Noncited Violation 05000482/2003004-02 and found that the licensee failed to correct a finding identified in NRC Inspection Report 05000482/2002008-01. A review of Off-Normal Procedure OFN RP-17, Control Room Evacuation after the licensee revised the procedure in response to a 2002 finding, revealed that the 2003 procedure revision had not corrected the problem, but had made the problem worse, by actually lengthening the allowed time to verify that a volume control tank outlet valve was closed, when it should have reduced the time allowed. If the valve was not closed in time, the centrifugal charging pumps could become gas bound and not pump water. On March 28, 2003, the licensee implemented a change to Procedure OFN RP-17 to ensure the valve was closed within the required time.
Example 6 - Multiple Failures to Correct Emergency Diesel Generator Heat Exchanger Tube Degradation The team reviewed the circumstance around NRC Inspection Report Noncited Violation 05000482/2002004-01 which closed URI 2002006-01: Failure to implement effective corrective actions for a significant condition adverse to quality for failing to perform eddy current testing of the Emergency Diesel Generator heat exchanger tubes until several tubes exhibited severe degradation. Specifically, on April 20, 1990, the licensee identified severe wall thinning of the Emergency Diesel Generator heat exchanger tubes requiring replacement, but did not implement corrective actions that recommended periodic eddy current evaluation of the Emergency Diesel Generator heat exchanger tubes to ensure continued structural integrity of the tubes. The licensee missed opportunities in 1993 and in 1997 to initiate periodic eddy current testing as well. As a result, as of December 13, 2001, the licensee had not performed eddy current examination of the Emergency Diesel Generator heat exchanger tubes until several tubes exhibited severe degradation.
d. Assessment of Safety-Conscious Work Environment
- (1) Inspection Scope The team interviewed more than 12 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-conscious work environment. The team also interviewed the site Employees Concern Program coordinator.
- (2) Assessment Of the individuals interviewed, one employee stated he was not aware of the licensees Employee Concern Program. A few of the others interviewed knew the program existed, but were not aware of who the program coordinator was or where his office was located. The inspection team observed a number of site bulletin boards and noted that none of the bulletin boards included information about the Employee Concerns Program. The Human Resources office stated that new employees were given information about the program and that refrigerator magnets containing information about the program had been issued to the employees in the past.
The team concluded that a positive safety-conscious work environment exists at Wolf Creek. The team determined that employees and contractors feel free to raise safety concerns to their supervision or bring concerns to the employees concern program. The team determined that licensee management is receptive to employee concerns and is willing to address issues raised by the latest safety culture survey.
e. Specific Issues Identified During This Inspection
- (1) Inspection Scope During this assessment the team performed the inspections scoped in Sections 4OA2 a.(1), 4OA2 b.(1), 4OA2 c.(1), and 4OA2 d.(1) above.
- (2) Findings and Observations Noncited Violation 05000482/2004006-01: Simulator Fidelity
Introduction.
A Green, self-revealing, noncited violation (NCV) was identified regarding simulator response to a transient condition. While completing immediate actions following a reactor trip which occurred on February 13, 2004, the Balance of Plant Operator observed what he understood to be a malfunction of the steam dump valves.
Subsequent investigation revealed that the plant systems operated properly but that the Balance of Plant Operator did not expect the steam generator atmospheric relief valves to be open while the steam dumps were closed shortly following a plant trip. The licensee identified that the simulator had not accurately modeled atmospheric relief valves post-trip operation since initial licensing.
Description.
Work Order 04-260586-000 was initiated to troubleshoot an apparent malfunction of the steam dump control system after a reactor trip on February 13, 2004.
The evaluation performed within the work order explained that the steam dumps operated properly and that the confusion was caused by the operation of the steam generator atmospheric relief valves. The Balance of Plant Operator did not expect the steam generator atmospheric relief valves to be open when the steam dump valves were closed. According to the Balance of Plant Operators understanding of steam generator atmospheric relief valves operation, the valves would not open until steam line pressure exceeded the steam generator atmospheric relief valves setpoint, which is approximately 1125 psi. However, due to the operation of the proportional-integral controller, the steam generator atmospheric relief valves may open below their setpoint and may stay open for a few minutes below the setpoint, well after steam dump valves have opened and reclosed.
The licensee identified that the confusion was caused by a failure to accurately model plant response in the simulator. According to the simulator model, the steam generator atmospheric relief valves would not open after most reactor trips. This simulator modeling error had been in place since initial licensing. A Simulator Change Request was issued on February 19, 2004.
Analysis.
This finding involved a licensed operator training deficiency regarding plant response to high power reactor trips. Therefore, this finding affected the Mitigating Systems Cornerstone since it impacted the operators response to mitigate the consequences of this transient and was considered more than minor since deficiencies in the operator training program could become a more significant safety concern if left uncorrected. Based on the results of a Significance Determination Process (SDP) using Manual Chapter (MC) 0609, Appendix I, this finding was determined to have very low safety significance, since it involved a simulator fidelity issue which impacted operator actions.
Enforcement.
Title 10 of the Code of Federal Regulations (CFR), Part 55.46(c),requires that plant referenced simulators used for operating tests or to meet experience requirements must demonstrate expected plant response to transient conditions to which the simulator was designed to respond. The Wolf Creek Nuclear Station simulator was designed to respond to reactor trips; however, the simulator response differed from actual plant response in that a normal reactor trip would cause the steam generator atmospheric relief valves to open, but not in the simulator. The failure to adequately model plant response in the simulator, discovered on February 19, 2004, is a violation of 10 CFR 55.46(c). This violation is being treated as a noncited violation 05000482/2004006-01 consistent with Section VI.A of the NRC Enforcement Policy.
Noncited Violation 05000482/2004006-02. Inadequate Design Control for Overcurrent Settings for Emergency Diesel Generator Supply Fan Breakers:
Introduction.
The team identified a 10 CFR 50, Appendix B, Criterion III noncited violation for failure to assure that design criteria had adequately been translated into specifications and procedures associated with the Emergency Diesel Generator supply fan breakers. In December 2002 and February, 2003, the Emergency Diesel Generator supply fan breakers had been replaced due to parts obsolescence. On March 12, 2003, Emergency Diesel Generator "A" supply fan Breaker NG03DBF6 was found tripped, but no problem was identified. On April 12 and April 15, 2003, additional failures of NG03DBF6 were identified. Evaluation determined that new breakers had been installed with overcurrent trips set too low to allow for the starting inrush current.
Description.
Emergency Diesel Generator "A" supply fan Breaker NG03DBF6 was replaced in December 2002 and Emergency Diesel Generator "B" supply fan Breaker NG04DBF6 was replaced in February, 2003. The previous breakers were ITE-Gould breakers and were replaced with Westinghouse/Cutler-Hammer breakers due to parts obsolescence. The replacement breakers tested satisfactorily and the modification was completed. The ITE-Gould breakers had been set at a trip setting of 2300 amps. The replacement breakers were set to a value of 2625 amps, the difference was to ensure that the replacement breaker did not trip spuriously due to the new style breakers wider tolerance.
On March 12, 2003, Breaker NG03DBF6 tripped open, and PIR 2003-0675 was written.
No specific problems were identified and the breaker was returned to service. On April 12, 2003, Breaker NG03DBF6 tripped again, and PIR 2003-1023 was initiated. A hardware problem was identified with a specific phase of the breaker causing the breaker to trip at a lower overcurrent value and the breaker was replaced.
On April 15, 2003, Breaker NG03DBF6 tripped opened again, and PIR 2003-1041 was initiated. This Performance Improvement Request identified that the fan motor inrush currents were well above the breakers instantaneous setting. During troubleshooting under Work Order 03-252082-000, the inrush currents seen during motor starts were well above the previously set instantaneous setting. The fan motor did not trip during the starts under the work order, even though the actual inrush currents were well above the instantaneous setting. The high inrush currents being well above the instantaneous setting of the breaker for 1/2 cycle or greater contributed to the spurious breaker trips.
The deficiency associated with this finding was the failure to adequately review the design requirements of the supply fans and incorporate these requirements into procedures and specifications . Specifically, on multiple occasions the licensee failed to identify and correct a concern with high inrush currents seen during the starting of the Emergency Diesel Generator supply fans. This failure potentially could affect the ability of the Emergency Diesel Generators to perform their design function during hot weather conditions. The supply fans are required to be operational when outside air temperatures are equal to or greater than 79 degrees F. During the time the new breakers were in service until the licensee identified and corrected the concern with the inrush current setting, the outside air temperatures had never exceeded 79 degrees F.
Analysis.
This finding is more than minor because it affected the mitigating systems cornerstone attribute of equipment performance and the cornerstone objective to ensure the availability of systems that respond to initiating events. This finding was evaluated using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 worksheet under the mitigating systems cornerstone, and was determined to be of very low safety significance because there was no actual loss of function, and the Emergency Diesel Generators were always operational.
Enforcement.
10 CFR 50, Appendix B, Criterion III, "Design Controls," states, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures and instructions. The failure to identify the required specifications for the inrush current for the new breakers and place this information in drawing, procedures or instructions, is a violation of 10 CFR 50, Appendix B, Criterion III. The failures of the Emergency Diesel Generator air supply fans occurred when the outside temperatures were less that 79 degrees F, and the supply fans were not required. As a result there was not actual loss of function because this violation is of very low safety significance and has been entered into the corrective action program as PIR 2003-1041, this violation is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy:
4OA6 Exit Meeting
The team discussed the findings with you and other members of the licensees staff on August 6, 2004 and again via telephone on September 29, 2004. Licensee management did not identify any materials examined during the inspection as proprietary.
4OA7 Licensee Identified Violations
The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositionsed a NCVs.
- Licensee Identified Noncited Violation: From 1984 through July 28, 2004, the licensee failed to follow its emergency plan designed to meet planning standard
- (5) in 10 CFR 50.47(b). On July 28, 2004 the licensee identified a violation of very low safety significance (Green), of 10 CFR 50.54(q) for failure to follow requirements of 10 CFR 50 Appendix E related to maintaining the offsite alert and notification system. Specifically, the licensee failed to provide tone alert radios to approximately 72 residences in areas of the Emergency Planning Zone (EPZ) where emergency siren sound levels were between 60 dB and 70 dB as committed to in the Federal Emergency Management Agency-approved Alert and Notification system. The licensee had committed in their alert and notification design report to distribute tone alert radios as the primary means of emergency notification to all occupied locations where siren sound levels were less than 70 db. For areas of low population density, such as the Wolf Creek emergency planning zone, FEMA REP-10 requires alternative means of notification, such as tone alert radios, in areas where the siren sound level is less than 60 db. Because the failure to ensure the distribution of tone alert radios in accordance with a licensee commitment to FEMA, and not a FEMA requirement, is of very low safety significance and has been entered into the licensees corrective action program (Performance Improvement Request 2004-1922), this violation is being treated as a Licensee Identified Non-Cited Violation, consistent with Section VI.A of the NRC Enforcement Policy.
- The licensee identified that they had failed to incorporate a preventative maintenance activity into a procedure due to closing a Performance Improvement Request corrective action to another Performance Improvement Request without ensuring the action was actually completed. Technical Specification 5.4.1 requires, in part, that written procedures be established, implemented, and maintained covering procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, section 9 includes "preventative maintenance schedules be developed to specify replacement of such items as filters. The failure to implement a PM program that included inspection and/or replacement of the Emergency Diesel Generator main air start distributor air filters is a violation of Technical Specification 5.4.1. Because this is a violation of very low safety significance and it has been entered into the corrective action program as PIR 2004-0132, this violation is being treated as a Licensee Identified Noncited Violation, consistent with Section VI.A of the NRC Enforcement Policy.
ATTACHMENT 1 PARTIAL LIST OF PERSONS CONTACTED Licensee T. Anselmi - Manager, Design Engineering P. Bedgood - Superintendent, Chemistry/Radiation Protection R. Calia - Manager, PI and L B. Dale - Acting Superintendent, Maintenance Support T. East - Superintendent, Emergency Planning D. Fehr - Manager, IS A. Harris - Director, PI and L P. Hawkins - Superintendent, Operations Support Work Control S. Hedges - Manager, Integrated Plant Scheduling D. Hooper - Supervisor, Licenseing S. Hopkins - Maintenance Support D. Jacobs - VP Operations and Plant Manager T. Jensen - Superintendent, Chemistry/Radiation Protection R. Kerving - Supervisor, Corrective Action Program M. Makar - Manager, Systems Engineering K. Moles - Manager, Regulatory Affairs K. Scherich - Director, Engineering C. Sibley - Regulatory Affairs M. Sunseri - VP Oversight M. Westman - Manager, Training J. Yunk - Manager, Organizational Performance Enclosure ITEMS OPENED AND CLOSED Opened and Closed 05000482/2004006-01 NCV Simulator Fidelity 05000482/2204006-02 NCV Inadequate design control for overcurrent settings for Emergency Diesel Generator supply fan breakers
DOCUMENTS REVIEWED
PLANT PROCEDURES
Procedure Title
AP 28A-001 Performance Improvement Request
AI 28A-011 Performance Improvement Request Initiation
AI 28A-012 Performance Improvement Request Screening
AI 28A-013 Performance Improvement Request Evaluation and Action
Plans
AI 28A-015 Performance Improvement Request Effectiveness Follow-
up
AI 28E-006 Common Cause Analysis
AI 28E-007 Performance Improvement Request Trending and Analysis
AP 28B-001 Root Cause Analysis
Desktop
Guidelines and Techniques for Meeting Management Expectations Regarding Significant
Performance Improvement Requests
-3-
Licensee Event Reports
PIR Number Title LER Number
2002-0048 Voluntary report of emergency diesel generator 2002-001-00
heat exchanger tube degradation
2002-1086 Mode Change with RCS Unidentified Leakage 2002-002-00
Greater than Technical Specification 3.4.13
2002-1180 Unit trip due to a feedwater regulating valve control 2002-003-00
card failure
2002-1898 Postulated fire event could lead to the loss of 2002-004-02
redundant trains of postfire safe shutdown
equipment
2002-2250 Engineering safety features actuation including 2002-005-00
emergency diesel generator start due to a
hardware failure in a relay driver card
2003-0010 Manipulation of component outside of procedural 2003-001-00
guidance causes reactor trip
2003-0805 Reactor vessel level indication system inoperable 2003-002-00
for period longer than allowed by Technical
Specifications
2003-2449 Reactor protection system actuation and reactor 2003-003-00
trip due to feedwater isolation valve closure
2003-3486 Failure of Safety Injection Accumulator Vent Line 2003-004-00
2004-0094 Inadequate Verification of Valve Position Following 2004-001
Testing Results in Technical Specification Violation
2004-0393 Reactor Protection System Actuation and Reactor 2004-002-00
Trip due to Main Feedwater Regulating Valve
Failing Closed
2004-0586 Automatic Start of B Emergency Diesel Generator 2004-003-00
due to Start-Up Transformer Cable Ground Fault
-4-
NONCITED VIOLATIONS REVIEWED
2002-0627 Transient Combustables 2001-006-01
2002-1247 Failed to follow procedure while drawing a vacuum 2002-002-01
on the RCS
2002-0048 Failure to implement appropriate corrective actions 2002-004-01
for degraded emergency diesel generator heat
exchanger tubes
2002-2393 Inadequate alternative shutdown procedure 2002-008-01
2003-0010 Manipulation of component outside of procedural 2003-003-01
guidance causes reactor trip
2003-0805,1713 Failure to ensure that emergency operating 2003-004-01
procedures could have been successfully
performed
2003-0333,0338 Failure to ensure that changes to an off-normal 2003-004-02
procedure were appropriate
2003-1553 Failure to critique an exercise performance 2003-004-03
deficiency relating to protecting nonessential
workers
2003-3220 Access control to radiologically significant areas 2003-006-02
2003-3069, 3135 ALARA planning and controls 2003-006-03
2003-1868 Failure to correctly translate a design basis into the 2003-007-01
internal flooding calculations for engineered safety
feature Switchgear Room 3302; thus the
assumptions used in Calculation FL-08 did not
agree with the as-built condition of the plant
2003-3704 Inadequate Fire Barriers at Seismic Gaps 2004-002-02
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NRC Information Notices:
IN TITLE PIRequest #
NRC INFORMATION NOTICE 2000-011 Licensee Responsibility for 2000-3328
Quality Assurance Oversight of Contractor Activities regarding
fabrication and use of Spent Fuel Storage Cask Systems
NRC INFORMATION NOTICE 2000-015 Recent Events Resulting in 2000-3480
Whole Body Exposures Excessive of Regulatory Limits
NRC INFORMATION NOTICE 2002-001 Metalclad Switchgear Failures 2002-1123
and ConsequentLosses of Offsite Power
NRC INFORMATION NOTICE 2002-002 Recent Experience with 2002-0316
Plugged Steam Generator Tubes
NRC INFORMATION NOTICE 2002-003 Highly Radioactive Particle 2002-0211
Control Problems During Spent Fuel Pool Cleanout
NRC INFORMATION NOTICE 2002-005 Foreign Material in Standby 2002-0564
Liquid Control Storage Tanks
NRC INFORMATION NOTICE 2002-013 Possible Indicators of 2002-0823
Ongoing Reactor Pressure Vessel Degradation
NRC INFORMATION NOTICE 2003-005 Failure to Detect Freespan 2002-3030
Cracks in PWR Steam Generator Tubes
NRC INFORMATION NOTICE 2003-011 Leakage Found on Bottom- 2003-1450
Mounted Instrumentation Nozzles
NRC INFORMATION NOTICE 2003-013 Steam Generator Tube 2003-1775
Degradation at Diablo Canyon
NRC INFORMATION NOTICE 2004-001 Auxiliary Feedwater Pump 2004-1224
Recirculation Line Orifice Fouling - Potential Common Cause Failure
PERFORMANCE IMPROVEMENT REQUESTS :
2000-1526 2002-0944 2003-2853 2003-1018 2004-1515 2004-0976
2000-3178 2002-0182 2003-2754 2003-0923 2004-1531 2004-0957
2001-0522 2002-2507 2003-2708 2003-0805 2004-1513 2004-0880
2001-1227 2002-3008 2003-3486 2003-0675 2004-1510 2004-0724
-6-
2001-1654 2002-0007 2003-3566 2003-0429 2004-1285 2004-0663
2001-1916 2002-0120 2003-2348 2003-0374 2004-1496 2004-0573
2001-1972 2002-0594 2003-1870 2003-0338 2004-1337 2004-0563
2001-1976 2002-1841 2003-1868 2003-0333 2004-1335 2004-0555
2001-1977 2002-1903 2003-1805 2003-0317 2004-1589 2004-0502
2001-1524 2002-2471 2003-1784 2003-0270 2004-1628 2004-0468
2002-2356 2002-2774 2003-1731 2003-0180 2004-1156 2004-0234
2002-2357 2003-3456 2003-1456 2003-0010 2004-1154 2004-0132
2002-2279 2003-3457 2003-1432 2003-0732 2004-1116 2004-0124
2002-1247 2003-3445 2003-1319 2003-1587 2004-1024 2004-0123
2002-0048 2003-3014 2003-1041 2003-1605 2004-0999 2004-0089
2002-1011 2003-2689 2003-1023 2003-1784 2004-0989 2004-0592
2002-1086 2003-3425 20021898 2003-3512 2004-1206
2002-1180 2002-1670 2002-2430 2003-3538
Level I and Level II PERFORMANCE IMPROVEMENT REQUESTS
2002-1394 2003-0010 2003-0688 2003-1683 2003-2883 2003-3792
2002-1401 2003-0078 2003-0805 2003-1952 2003-3069 2004-0094
2002-1434 2003-0140 2003-0874 2003-1976 2003-3072 2004-0149
2002-1472 2003-0145 2003-0890 2003-2372 2003-3085 2004-0393
2002-1575 2003-0207 2003-0903 2003-2438 2003-3136 2004-0496
2002-1582 2003-0218 2003-1199 2003-2449 2003-3201 2004-0586
2002-1704 2003-0439 2003-1255 2003-2468 2003-3220 2004-0909
2002-1834 2003-0440 2003-1259 2003-2502 2003-3473 2004-1101
2002-1845 2003-0529 2003-1431 2003-2600 2003-3486 2004-1505
2002-2087 2003-0569 2003-1595 2003-2698 2003-3513
2002-2250 2003-0587 2003-1647 2003-2741 2003-3768
-7-
OTHER
Corrective Action Review Board Charter
Root Cause Manual
Trending Code Manual
Level I and II Performance Improvement Requests Generated June
2002 - June 2004
Level I and II Performance Improvement Requests Closed June
2002 - June 2004
Temporary Modification Order (TMO) 04-003-KJ: Temporary Clamp 03/08/2004
to Prevent Fuel Supply Header Return Line from Pulling out of
Tubing Connection
Operability Evaluation (OE) KJ-04-002: B Emergency Diesel Revision 0
Generator Failed to Achieve the 12 Second Start Time Requirement
Root Cause Analysis Report, PIR 2003-0805, Inoperability of 4/30/03
Forced Flow Reactor Vessel Level Indication
Root Cause Analysis Report, PIR 2003-3486, Socket Weld Failure 03/28/04
of EPV 0109 Vent Line
Cause Determination, PIR 2003-0010, for reactor trip on 1/3/03 9/12/03
cause by loss of both control rod drive motor generator output
Cause Determination, PIR 2002-0048, for potential common mode 6/24/03
failure resulting from degraded tubes in the Emergency Diesel
Generators heat exchangers
Enclosure