IR 05000315/2004010
| ML042880244 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/13/2004 |
| From: | Eric Duncan NRC/RGN-III/DRP/RPB6 |
| To: | Nazar M American Electric Power Co |
| References | |
| IR-04-010 | |
| Download: ML042880244 (43) | |
Text
ber 13, 2004
SUBJECT:
D. C. COOK NUCLEAR POWER PLANT, UNITS 1 AND 2 NRC INTEGRATED INSPECTION REPORT 05000315/2004010; 05000316/2004010
Dear Mr. Nazar:
On September 30, 2004, the U. S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your D. C. Cook Nuclear Power Plant, Units 1 and 2. The enclosed report documents the inspection findings which were discussed on September 30, 2004, with Mr. J. Jensen and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Based on the results of this inspection, no findings of significance were identified. A licensee-identified violation which was determined to be of very low safety significance is discussed in Section 4OA7 of this report.
If you contest the subject or severity of a Non-Cited Violation, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the U. S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the D. C. Cook Nuclear Power Plant. 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 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/
Eric R. Duncan, Chief Branch 6 Division of Reactor Projects Docket Nos. 50-315; 50-316 License Nos. DPR-58; DPR-74
Enclosure:
Inspection Report 05000315/2004010; 05000316/2004010 w/Attachment: Supplemental Information
REGION III==
Docket Nos: 50-315; 50-316 License Nos: DPR-58; DPR-74 Report No: 05000315/20040010; 05000316/2004010 Licensee: Indiana Michigan Power Company Facility: D. C. Cook Nuclear Power Plant, Units 1 and 2 Location: 1 Cook Place Bridgman, MI 49106 Dates: July 1, 2004, through September 30, 2004 Inspectors: B. Kemker, Senior Resident Inspector I. Netzel, Resident Inspector A. Dunlop, Senior Reactor Engineer B. Jose, Reactor Engineer R. Lerch, Senior Project Engineer P. Lougheed, Senior Reactor Engineer R. Ruiz, Reactor Engineer S. Sheldon, Senior Reactor Engineer W. Slawinski, Senior Radiation Specialist Approved by: Eric R. Duncan, Chief Branch 6 Division of Reactor Projects Enclosure
SUMMARY OF FINDINGS
IR 05000315/2004010, IR 05000316/20040010; 07/01/2004-09/30/2004; D. C. Cook Nuclear
Power Plant, Units 1 and 2; Integrated Inspection Report.
This report covers a 13-week period of inspection by resident and region-based inspectors. No findings of significance were identified. 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.
A. Inspector-Identified and Self-Revealed Findings No findings of significance were identified.
Licensee Identified Violations
A violation of very low safety significance, which was identified by the licensee, has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. The violation and the licensees corrective action tracking number is listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
Unit 1 and Unit 2 were operated at or near full power during the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04 Equipment Alignment
.1 Partial System Walkdowns
a. Inspection Scope
The inspectors performed three partial system walkdowns of the following risk significant systems:
C Unit 2 Containment Spray System C Unit 2 East and West Auxiliary Feedwater Systems C Unit 1 West Centrifugal Charging System The inspectors selected these systems based on their risk significance relative to the reactor safety cornerstones. The inspectors reviewed operating procedures, system diagrams, Technical Specification (TS) requirements, Administrative TSs, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also walked down accessible portions of the systems to verify system components were aligned correctly.
In addition, the inspectors verified that equipment alignment problems were entered into the licensees corrective action program with the appropriate characterization and significance.
b. Findings
No findings of significance were identified.
.2 Complete System Walkdown
a. Inspection Scope
The inspectors performed one complete system walkdown of the following risk significant system:
- Unit 1 Component Cooling Water System performed from July 29, 2004 through August 20, 2004 The inspectors reviewed ongoing system maintenance, open job orders, and design issues for potential effects on the ability of the system to perform its design functions.
The inspectors reviewed operating procedures, system diagrams, TS requirements, and applicable sections of the Updated Final Safety Analysis Report (UFSAR) to ensure the correct system lineup. The inspectors verified acceptable material condition of system components, availability of electrical power to system components, and that ancillary equipment or debris did not interfere with system performance.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
.1 Routine Resident Inspector Tours
a. Inspection Scope
The inspectors performed eight fire protection walkdowns of the following risk significant plant areas:
C Unit 1 West Non-Essential Service Water Valve Area (Zone 33B)
C Unit 2 West Non-Essential Service Water Valve Area (Zone 34B)
C Units 1 and 2 Auxiliary Building North - Elevation 609' (Zone 44N)
C Units 1 and 2 Auxiliary Building South - Elevation 609' (Zone 44S)
C Unit 1 Auxiliary Cable Vault - Elevation 620'6" (Zone 56)
C Unit 2 Auxiliary Cable Vault - Elevation 620'6" (Zone 59)
C Unit 1 Charging Pump Rooms (Zone 62)
C Unit 2 Charging Pump Rooms (Zone 63)
The inspectors verified that fire zone conditions were consistent with assumptions in the licensee's Fire Hazards
Analysis.
The inspectors walked down fire detection and suppression equipment, assessed the material condition of fire fighting equipment, and evaluated the control of transient combustible materials. In addition, the inspectors verified that fire protection related problems were entered into the licensee's corrective action program with the appropriate characterization and significance.
b. Findings
No findings of significance were identified.
.2 Annual Fire Drill Observation
a. Inspection Scope
The inspectors assessed fire brigade performance and the drill evaluators critique during a fire brigade drill conducted in the Unit 1 west essential service water pump room on August 18, 2004. The drill simulated an electrical fire in the pumps motor. The inspectors focused on command and control of fire brigade activities, fire fighting and communication practices, material condition and use of fire fighting equipment, and implementation of pre-fire plan strategies.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures
a. Inspection Scope
The inspectors performed one inspection activity related to the licensees precautions to mitigate the risk from internal flooding events. The following inspection activities were performed:
C the inspectors reviewed the Unit 1 and Unit 2 Flooding Evaluation reports, the UFSAR and other selected design basis documents to identify those areas susceptible to internal flooding; C the inspectors performed a walkdown of the lower elevations of the Turbine Building and Auxiliary Building to assess the adequacy of watertight doors and verify that drains and sumps were clear of debris and were operable; and C the inspectors reviewed selected operating procedures used to identify and mitigate flooding events and verified that these procedures were adequate.
C the inspectors reviewed the licensees inspection of underground manholes susceptible to external flooding which contained risk-significant cables and verified that these inspections were adequate.
In addition, the inspectors verified that flood protection related problems were entered into the licensee's corrective action program with the appropriate characterization and significance.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification
.1 Resident Inspector Quarterly Review
a. Inspection Scope
The inspectors assessed licensed operator performance and the training evaluators critique during a licensed operator requalification evaluation in the D. C. Cook Plant operations training simulator on July 20, 2004. The inspectors focused on alarm response, command and control of crew activities, communication practices, procedural adherence, and implementation of emergency plan requirements.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
.1 Resident Inspector Quarterly Review
a. Inspection Scope
The inspectors evaluated the licensees handling of selected degraded performance issues involving the following two risk-significant structures, systems, and components (SSCs):
C Unit 1 #12 Reactor Coolant Pump #1 Seal Erratic Leakoff C Unit 1 and 2 Post Accident Containment Hydrogen Monitoring System Backup Air Pressure Regulator Failures The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the SSCs. Specifically, the inspectors independently verified the licensees handling of SSC performance or condition problems in terms of:
C appropriate work practices, C identifying and addressing common cause failures, C scoping of SSCs in accordance with 10 CFR 50.65(b),
C characterizing SSC reliability issues, C tracking SSC unavailability, C trending key parameters (condition monitoring),
C 10 CFR 50.65(a)(1) or (a)(2) classification and reclassification, and C appropriateness of performance criteria for SSCs/functions classified (a)(2)and/or appropriateness and adequacy of goals and corrective actions for SSCs/functions classified (a)(1).
In addition, the inspectors verified that problems associated with the effectiveness of plant maintenance were entered into the licensees corrective action program with the appropriate characterization and significance.
b. Findings
No findings of significance were identified.
.2 Periodic Evaluation
a. Inspection Scope
The inspectors reviewed the Maintenance Rule periodic evaluation report, which was required per 10 CFR 50.65(a)(3). This evaluation was a periodic assessment of the effectiveness of maintenance for those SSCs included within the scope of the rule. The licensee monitors SSCs where maintenance has not been effective, by either excessive failures or unavailability, under (a)(1) of the rule, such that the SSCs receive appropriate attention to correct the deficiencies. The licensee monitors the remaining SSCs where maintenance has been demonstrated as being effective under (a)(2) of the rule, to ensure the SSCs will continue to be able to perform their intended function. The objectives of the inspection were to:
- Verify that the periodic evaluation was completed within the time constraints defined in 10 CFR 50.65 (once per refueling cycle, not to exceed 2 years),ensuring that the licensee reviewed its goals, monitoring, preventive maintenance activities, industry operating experience, and made appropriate adjustments as a result of that review;
- Verify that the licensee balanced reliability and unavailability for safety significant SSCs during the previous refueling cycle;
- Verify for SSCs being monitored under (a)(1) of the rule, that goals were being met, corrective actions were appropriate to correct the defective condition including the use of industry operating experience, and (a)(1) activities and related goals were adjusted as needed; and
- Verify that the licensee had established (a)(2) performance criteria, examined any SSCs that failed to meet the performance criteria, or reviewed any SSCs that had incurred repeated maintenance preventable functional failures, including a verification that the SSCs were considered for monitoring under (a)(1) of the rule.
The inspectors examined the periodic evaluation report for the time frame of October 5, 2001 through June 30, 2003. To evaluate the effectiveness of (a)(1) and (a)(2)activities, the inspectors examined (a)(1) action plans, justifications for returning SSCs from (a)(1) to (a)(2), and a number of condition reports to evaluate the licensees functional failure determinations. In addition, the condition reports were reviewed to verify that the threshold for identification of problems was at an appropriate level and the associated corrective actions were appropriate. The inspectors focused the inspection on the following systems (5 samples):
- Essential Service Water
- Emergency Diesel Generator Ventilation
- 250 Volt Direct Current (DC) Power
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for the following six maintenance and operational activities affecting safety-related equipment:
C Unit 1 Ice Condenser Cooling Relief Valve Lifting C Unit 1 CD Emergency Diesel Generator Starting Air Actuator Replacement Emergent Work C Unit 2 Turbine Driven Auxiliary Feedwater Pump Maintenance C Unit 1 Pressurizer Sensing Line Root Valve Leak C Unit 2 Switchyard High Risk Maintenance Activities C Unit 2-IMO-331 Residual Heat Removal Spray Valve Planned Maintenance These activities were selected based on their potential risk significance relative to the reactor safety cornerstones.
As applicable for each of the above activities, the inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst and/or shift technical advisor, and verified that plant conditions were consistent with the risk assessment assumptions. The inspectors also reviewed TS requirements and walked down portions of redundant safety systems, when applicable, to verify that risk analysis assumptions were valid and applicable requirements were met.
In addition, the inspectors verified that maintenance risk-related problems were entered into the licensee's corrective action program with the appropriate characterization and significance.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed the following condition reports (CRs) to ensure that either the condition did not render the involved equipment inoperable or result in an unrecognized increase in plant risk, or the licensee appropriately applied TS limitations and appropriately returned the affected equipment to an operable status.
C CR P-99-07602, "Momentary Ratings Exceeded on 4KV Breaker for Fault Conditions" C CR P-99-08330, "Calculation Shows Fault Currents at the 4KV Buses are Higher than Manufacturers Rating" C CR 03099017, "TS 3.6.3.1 Applicability With Respect to Certain Dual Function Valves," and CR 03098003, "2-ICM-250 Failed to Stroke Open Within Maximum Stroke Time" C CR 04208040, "2-GFW-L-825 Material Loss Due to an Aggressive Environment" C CR 04061032, "Undersized O-Ring Installed on 2-PP-50W West Centrifugal Charging Pump" C CR 04165014, "Performance of 2-OHP-4021-055-005 Feed Pump Turbine Miscellaneous Trip Test Causes Both Motor Driven Auxiliary Feedwater Pumps to Become Inoperable" C CR 04175023, "No Testing Is Performed on Slave Relay K627, Reactor Coolant Pump Trip on Bus Under-frequency Output Relays" C CR 01125013, "#4 Accumulator Filled When the South Safety Injection Pump Was Started" In addition, the inspectors verified that problems related to the operability of safety-related plant equipment were entered into the licensees corrective action program with the appropriate characterization and significance.
b. Findings
No findings of significance were identified.
1R19 Post Maintenance Testing
a. Inspection Scope
The inspectors reviewed six post maintenance testing activities associated with the following scheduled maintenance:
C Unit 1 Turbine Driven Auxiliary Feedwater Pump Planned Maintenance C Unit 1 AB Emergency Diesel Generator Planned Maintenance C Unit 1 East Essential Service Water Pump Discharge Valve (1-WMO-701)
Planned Maintenance C Unit 2 South Safety Injection Pump Planned Maintenance C Unit 2 West Motor Driven Auxiliary Feedwater Pump Recirculating Line Check Valve Replacement Planned Maintenance C Unit 2 Reactor Protection Control Group Cabinet #17 Math Unit Replacement Emergent Repair The inspectors reviewed the scope of the work performed and evaluated the adequacy of the specified post maintenance testing. The inspectors verified that the post maintenance testing was performed in accordance with approved procedures, that the procedures clearly stated acceptance criteria, and that the acceptance criteria were met.
The inspectors interviewed operations, maintenance, and engineering department personnel and reviewed the completed post maintenance testing documentation.
In addition, the inspectors verified that post maintenance testing problems were entered into the licensees corrective action program with the appropriate characterization and significance.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed portions of the following five surveillance testing activities and/or reviewed the test results to determine whether risk significant systems and equipment were capable of performing their intended safety function and to verify that testing was conducted in accordance with applicable procedural and TS requirements.
C Unit 1 East Component Cooling Water System Test C Unit 1 AB Emergency Diesel Generator Monthly Test C Unit 1 "B" and "C" Local Leak Rate Test, Penetration 82 C Unit 2 West Essential Service Water System Test C Unit 2 "B" and "C" Local Leak Rate Test, Penetration 82 The inspectors reviewed the test methodology and test results to verify that equipment performance was consistent with safety analysis and design basis assumptions. In addition, the inspectors verified that surveillance testing problems were being entered into the corrective action program with the appropriate significance characterization.
The inspectors also reviewed the last containment local leak rate test results for penetration 82 in both Unit 1 and Unit 2. This included an overall review of the procedure, review of specific work packages for four containment isolation valves, and review of the method used to verify compliance with the TS for maintaining total leakage from all containment penetrations within 60 percent of the allowable leakage. Because both the inner and outer valves were tested during the same surveillance, the inspectors considered review of both the inboard and outboard valves on one unit to be one sample under the baseline inspection program.
b. Findings
No findings of significance were identified.
1R23 Temporary Modifications
a. Inspection Scope
The inspectors reviewed two temporary modifications and verified that the installation was consistent with design modification documents and that the modifications did not adversely impact system operability or availability.
- 1-TM-04-46-R0, "Instrumentation to Monitor Vibration Levels on Unit 1 Circulating Water Pump Motors"
- ICP-01024, "Unit 1 Pressurizer Enclosure High Temperature Alarm Change,"
and ICP-01027, "Unit 1 Lower Containment Temperature Alarm Setpoint Change for 1-SG-18 Recorder Point Numbers 8, 11, and 12" The inspectors verified that configuration control of the modifications were correct by reviewing design modification documents and confirmed that appropriate post-installation testing was accomplished. The inspectors interviewed engineering, and operations department personnel and reviewed the design modification documents and 10 CFR 50.59 evaluations against the applicable portions of the TS and UFSAR.
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
The inspectors observed activities in the plant simulator, the Operations Support Center, and the Technical Support Center during an emergency preparedness training drill conducted on July 27, 2004. The inspectors verified that the emergency classifications and notifications to offsite agencies were completed in an accurate and timely manner as required by the emergency plan implementing procedures. The inspectors also verified that the training drill was conducted in accordance with the prescribed sequence of events, drill objectives were satisfied and that the required prompts from the licensee drill controllers were appropriately communicated to the drill participants.
The inspectors observed the post-drill critique in the Technical Support Center and reviewed documented post-drill critique comments by licensee evaluators to verify that licensee personnel and licensee drill evaluators adequately self-identified drill performance problems of significance. The inspectors also verified that condition reports were generated for drill performance problems of significance and entered into the corrective action program with the appropriate characterization and significance.
b. Findings
No findings of significance identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 Review of Licensee Performance Indicators for the Occupational Exposure Cornerstone
a. Inspection Scope
The inspectors reviewed licensee event reports, corrective action documents and data reported on the NRCs web site relative to the licensees occupational exposure control performance indicator to determine whether or not the conditions surrounding any actual or potential performance indicator occurrences had been evaluated, and identified problems had been entered into the corrective action program for resolution.
This review represented one inspection sample.
b. Findings
No findings of significance were identified.
.2 Plant Walkdowns/Boundary Verifications and Radiation Work Permit Reviews
a. Inspection Scope
The inspectors identified work areas during the inspection located within high and locked high radiation areas of the plant and selectively reviewed radiation work permit (RWP) packages and radiation surveys for these areas. The inspectors evaluated the radiological controls for these areas to determine if these controls including postings and access control barriers were adequate.
The inspectors reviewed active RWPs and as-low-as-is-reasonably-achievable (ALARA)plans which governed activities in radiologically significant areas to identify the work control instructions and control barriers that had been specified. Electronic dosimeter alarm set points for both integrated dose and dose rate were evaluated for conformity with survey indications and RWP policy. Workers were questioned by the inspectors to verify that they were aware of the actions required when their electronic dosimeters malfunctioned or alarmed.
The inspectors walked down and surveyed (using an NRC survey meter) radiologically significant area boundaries in the Unit 1 and 2 Auxiliary Building and at the fuel transfer tube access barriers in the Unit 1 and 2 Turbine Building to verify that the necessary radiological access controls were in place, that licensee postings were complete and accurate, and that physical barricades/barriers were adequate. During the walkdowns, the inspectors challenged access control boundaries to verify that high radiation area, locked high radiation area (LHRA), and very high radiation area (VHRA) access was controlled consistent with the licensees procedures, TSs, the requirements of 10 CFR 20.1601 and 20.1602 and were consistent with Regulatory Guide 8.38, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants."
The inspectors reviewed RWP and post job review documents for selected activities performed in 2004 to verify barrier integrity and engineering controls performance (e.g., filtered ventilation system operation) and to determine if there was a potential for individual worker internal exposures of >50 millirem committed effective dose equivalent. The inspectors reviewed the licensees procedures and its methods for the assessment of internal dose as required by 10 CFR 20.1204, to ensure methodologies were technically sound and included an assessment of the impact of hard to detect radionuclides such as pure beta and alpha emitters, as applicable. No worker intakes occurred since internal dose calculations were last reviewed by the inspectors as described in NRC Inspection Report 50-315/03-16(DRS); 50-316/03-16(DRS).
The inspectors reviewed the licensees physical and programmatic controls for highly activated and/or contaminated materials (non-fuel) stored within spent fuel or other storage pools. To conduct this review, radiation protection (RP) staff were interviewed, RP and foreign material exclusion procedures were reviewed as was the latest inventory record for the spent fuel pool and transfer canal. Given that the licensee seldom stores activated or highly contaminated materials in the spent fuel pool in a manner which would allow them to be readily removed, the radiological controls for such storage was discussed with RP supervisors to ensure adequate barriers would be established to reduce the potential for the inadvertent movement of irradiated materials.
These reviews represented five inspection samples.
b. Findings
No findings of significance were identified.
.3 Problem Identification and Resolution
a. Inspection Scope
The inspectors reviewed the results of a recently completed self-assessment of the radiological access control program, the condition report database for all RP-related issues generated between January and July 2004 along with individual condition reports related to the radiological access and exposure control programs to verify that identified problems were entered into the corrective action program for resolution. The inspectors screened all high radiation area (HRA) related radiological incidents and reviewed those which were of greatest potential significance (non-performance indicator occurrences identified by the licensee in high and locked high radiation areas) to verify that follow-up activities were conducted in an effective and timely manner commensurate with their importance to safety and risk based on the following:
- Initial problem identification, characterization, and tracking;
- Disposition of operability/reportability issues;
- Evaluation of safety significance/risk and priority for resolution;
- Identification of repetitive problems;
- Identification of contributing causes; and
- Identification and implementation of corrective actions.
The inspectors evaluated the licensees process for problem identification, characterization, prioritization, and verified that problems were entered into the corrective action program and were being resolved. For repetitive deficiencies, the inspectors verified that the licensees self-assessment activities were capable of identifying and addressing these deficiencies, if applicable.
The inspectors reviewed licensee documentation packages for all potential performance indicator events occurring since the last inspection to determine if any of these events involved dose rates >25 Rem/hr at 30 centimeters or >500 Rem/hr at 1 meter or involved unintended exposures >100 millirem total effective dose equivalent (or >5 Rem shallow dose equivalent or >1.5 Rem lens dose equivalent). None were identified.
These reviews represented four inspection samples. Specifically, the samples pertained to the licensees self-assessment capabilities, its problem identification and resolution program for HRA incidents, a review of the licensees ability to identify and address repetitive deficiencies, and a review of those potential performance indicator occurrences of greatest radiological risk.
b. Findings
No findings of significance were identified.
.4 Job-In-Progress Reviews and Review of Work Practices in Radiologically Significant
Areas
a. Inspection Scope
The inspectors accompanied RP and maintenance staffs into the Unit 2 boric acid storage tank room and evaluated filter change-out activities and the associated radiological controls. The inspectors reviewed the radiation survey and associated radiological job requirements for this work activity as provided in the RWP package, and attended the pre-job briefing to assess the adequacy of the information exchanged. The inspectors also reviewed the RWP and ALARA plan and attended the pre-job briefing for the handling of high level radioactive waste (filters) in the drumming room, an activity scheduled to take place the week following the inspection.
Job performance during the boric acid filter replacement was observed to verify that radiological conditions in the work area were adequately communicated to workers through the pre-job brief and postings. The inspectors also verified the adequacy of the radiological oversight provided by the radiation protection staff including the radiological surveys and radiation protection technician job coverage.
Previously completed work in high radiation areas that had significant dose rate gradients were reviewed to evaluate the application of dosimetry to effectively monitor exposure to personnel. Also, the inspectors reviewed the licensees procedure and its generic practices associated with dosimetry placement, use of extremity dosimetry, practices for monitoring neutron exposure and for the use of multiple whole body dosimetry for work in areas with significant dose rate gradients for compliance with the requirements of 10 CFR 20.1201(c) and applicable industry guidelines.
The inspectors also reviewed the licensees procedures and discussed with RP staff its practices for at-power containment entries and for entry into the reactor pit and in-core detector instrument room to determine the adequacy of the radiological controls and hazards assessment associated with such entries. Work instructions provided in radiation work permits, pre-entry briefings and the administrative and physical controls to prevent unauthorized entry into these radiologically significant areas were discussed with RP staff to determine their adequacy relative to industry practices and NRC Information Notices.
These reviews represented three inspection samples.
b. Findings
No findings of significance were identified.
.5 High Risk Significant, LHRA and VHRA Access Controls
a. Inspection Scope
The inspectors reviewed the licensees procedures, associated RP guidelines and evaluated RP practices for the control of access to radiologically significant areas (high, locked high, and very high radiation areas), and assessed compliance with the licensees TSs, the requirements of 10 CFR Part 20, and the guidance contained in Regulatory Guide 8.38. In particular, the inspectors evaluated the RP staffs control of keys to LHRAs and VHRAs, the use of RP supervisors to control access into these areas while work is taking place, and methods and practices for independently verifying proper closure and locking of access doors upon area egress. The inspectors selectively reviewed high radiation area key issuance/return and inventory records for the second quarter of 2004 to verify the adequacy of accountability practices and documentation. The inspectors also reviewed the licensees procedure and practices for radiation protection manager and plant ALARA Committee approval for access into VHRAs for compliance with 10 CFR 20.1602.
The inspectors discussed with RP staff the controls that were in place for areas that had the potential to become high radiation areas or greater during certain plant operations to determine if these plant operations required communication before hand with the RP group, so as to allow corresponding timely actions to properly post and control the radiation hazards. In particular, operations procedures for reactor coolant drain tank and volume control tank operations, for residual heat removal system actuation and for other reactor operations activities that could affect plant radiological conditions were reviewed along with recently developed radiological survey guidance to determine if adequate mechanisms were in-place to identify and control potential emerging radiological hazards.
The inspectors conducted plant walkdowns to verify the adequacy of postings and physical barriers including the locking of entrances to numerous LHRAs and those VHRAs accessible to the inspectors.
These reviews represented three inspection samples.
b. Findings
No findings of significance were identified.
.6 Radiation Worker Performance
a. Inspection Scope
During the filter replacement work in the boric acid storage tank room, the inspectors evaluated radiation worker performance with respect to stated radiation protection work requirements and to determine whether workers were aware of the radiological conditions, the RWP controls and limits in place, and that their performance had accounted for the level of radiological hazards present.
The inspectors reviewed radiological problem reports generated from January 2004 through July 2004 which found that the cause or contributor to the event was radiation worker error to determine if there was an observable pattern traceable to a similar cause, and to determine if this matched the corrective action approach taken by the licensee to resolve the identified problems.
These reviews represented two inspection samples.
b. Findings
No findings of significance were identified.
.7 Radiation Protection Technician Proficiency
a. Inspection Scope
During job observations, the inspectors evaluated radiation protection technician performance with respect to radiation protection work requirements, conformance with procedures and those requirements specified in the RWP, and to determine if their performance was consistent with the radiological hazards that existed.
The inspectors reviewed selected radiological problem reports generated between December 2003 and August 2004 to determine the extent of any specific problems or trends caused by RP technician errors, and to determine if the corrective action approach taken by the licensee to resolve the reported problems, if applicable, was adequate.
These reviews represented two inspection samples.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification
Cornerstones: Mitigating Systems and Barrier Integrity
.1 Safety System Functional Failures
a. Inspection Scope
The inspectors verified the Safety System Functional Failures Performance Indicator for both units. The inspectors reviewed each Licensee Event Report (LER) from October 2003 to June 2004, determined the number of safety system functional failures that occurred, evaluated each LER against the performance indicator definitions, and verified the number of safety system functional failures reported.
b. Findings
No findings of significance were identified.
.2 Reactor Coolant System Leakage
a. Inspection Scope
The inspectors verified the Reactor Coolant System (RCS) Leakage performance Indicator for both units. The inspectors reviewed operating logs and the results of RCS water inventory balance calculations performed from October 2003 through June 2004 and verified the licensees calculation of RCS leakage for both units.
b. Findings
No findings of significance were identified.
.3 Reactor Coolant System Specific Activity
a. Inspection Scope
The inspectors verified the Reactor Coolant System Specific Activity performance Indicator for both units. The inspectors reviewed chemistry department records including selected isotopic analyses for the period April 2003 through mid-August 2004, to verify that the greatest Dose Equivalent Iodine (DEI) values determined during steady state operations for those months corresponded with the values reported to the NRC.
The inspectors selectively reviewed DEI calculations to verify that the appropriate conversion factors were used in the assessment as required by the licensees procedure. Also, sample analyses and DEI calculation methods were discussed with chemistry staff to determine their adequacy. To verify the accuracy of the performance indicator data reported, performance indicator definitions and guidance contained in Revision 2 of Nuclear Energy Institute 99-02, "Regulatory Assessment Performance Indicator Guideline," were used.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
.1 Routine Review of Identification and Resolution of Problems
a. Inspection Scope
As discussed in previous sections of this report, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that they were being entered into the licensee's corrective action system at an appropriate threshold, that adequate attention was being given to timely corrective actions, and that adverse trends were identified and addressed. Some minor issues entered into the licensee's corrective action system as a result of inspectors' observations are included in the list of documents reviewed which are attached to this report.
b. Findings
No findings of significance were identified.
.2 Annual Sample Review
a. Inspection Scope
The inspectors selected the following two issues for in-depth review:
- CR 03016039, "Significant Operating Event Report 03-01, Emergency Power Reliability"
- CR 04128084, "Troubleshooting Steps Delineated in Job Order 04001006-01 Only Partially Covered the Vendor Recommended Troubleshooting Steps" The inspectors verified the following attributes during their review of the licensee's corrective actions for the above condition reports and other related condition reports:
- consideration of the extent of condition, generic implications, common cause and previous occurrences;
- classification and prioritization of the resolution of the problem, commensurate with safety significance;
- identification of the root and contributing causes of the problem; and
- identification of corrective actions which were appropriately focused to correct the problem.
The inspectors discussed the corrective actions and associated condition report evaluations with site personnel.
b. Findings
No findings of significance were identified. However, the inspectors had the following observations regarding the licensee's troubleshooting efforts, condition evaluation and corrective actions associated with CR 04128084 and other related condition reports.
During a Unit 2 reactor startup on January 1, 2004, control rod K-10 dropped 60 steps.
The licensee formed a troubleshooting team to address the problem and evaluated the condition. The inspectors provided several observations at that time and several condition reports were generated as a result. Those observations included:
1) the troubleshooting procedure for measuring the resistance of the gripper coil connector used the wrong acceptance criteria (CR 04128083);2) not all of the applicable troubleshooting steps provided in the vendor manual were used (CR 04128084); and 3) vendor identified deficiencies in previously performed job orders had not been evaluated (CR 04128085).
The inspectors found that the condition report evaluation for CR 04128084 was ambiguous and did not appear to acknowledge that mistakes had been made during the licensee's troubleshooting. The evaluations justified the actions of the troubleshooting team, provided unrelated information, and did not reference plant procedures which would have more succinctly answered the inspectors' questions. Resolution of these concerns required additional inspector involvement and the condition report evaluations were revised appropriately.
As a result, the inspectors concluded that replacing the K-10 gripper coil connector was unnecessary, that two other connectors were damaged in the process of replacing the K-10 connector, and that the probable cause of the rod slippage was not determined although the problem did not recur during the subsequent startup.
4OA3 Event Response
.1 (Closed) LER 50-315/2004-001-00: "Failure to Comply with TS 3.7.5.1, Control Room
Emergency Ventilation System." The licensee identified that combined leakage from the Unit 2 South safety injection pump discharge valve (2-SI-206) manual isolation pressure seal ring and seat leakage past the pump's three drain valves was greater than that assumed in the safety analysis for TS 3.7.5.1. This condition existed for 66 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br /> and 43 minutes from July 18, 2003 through July 21, 2003, and rendered the Control Room emergency ventilation system filter units for both Unit 1 and Unit 2 inoperable. The licensee failed to recognize at that time that this condition constituted a failure to meet the requirements of TS 3.7.5.1 and failed to comply with the 24-hour action requirement to restore the Control Room emergency ventilation system filter unit to an operable status or be in at least Mode 3 (Hot Standby) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 (Cold Shutdown) within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The licensee discovered this non-compliance on May 4, 2004, while investigating the cause of a very slow decrease in volume from the Unit 2 refueling water storage tank, which revealed the problem with the leaking drain valves. The licensee reported this event as a condition which was prohibited by the plants TSs in accordance with 10 CFR 50.73(a)(2)(i)(B). The licensee determined that the apparent cause for this event was excessive leakage from 2-SI-206 and the pumps three drain valves. In addition, the licensee identified that its RCS leakage monitoring program did not include an accounting of seat leakage past numerous emergency core cooling and containment spray system pump casing and nozzle drain lines that would be relied upon to isolate the drain lines from the rest of the system piping that must remain in service during the recirculation mode following an accident.
The inspectors concluded that the licensee implemented reasonable immediate corrective actions for this event including replacing the pumps three drain valves and repairing 2-SI-206. The inspectors concluded that this event was a licensee performance deficiency warranting a significance evaluation.
The inspectors assessed this finding using the Significance Determination Process (SDP). The inspectors reviewed the samples of minor issues in NRC Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," and determined that there were no examples related to this issue.
Consistent with the guidance in IMC 0612, "Power Reactor Inspection Reports,"
Appendix B, "Issue Disposition Screening," the inspectors determined that this failure to comply with the TS Limiting Condition for Operation action requirement could become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly implement the requirements of the TS 3.7.5.1 could reasonably result in conditions affecting habitability for Control Room operators during an accident. Because this issue affected the operability of the Control Room emergency ventilation system, the inspectors concluded that this issue was associated with the barrier integrity cornerstone. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations,"
and determined that this finding was a licensee performance deficiency of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the Control Room.
The inspectors concluded that this event constituted a violation of TS 3.7.5.1. A licensee-identified Non-Cited Violation is documented in Section 4OA7.1 of this report.
This LER is closed.
4OA6 Meetings
.1 Resident Inspectors Exit Meeting
The inspectors presented the inspection results to Mr. J. Jensen and other members of licensee management at the conclusion of the inspection on September 30, 2004. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.
Proprietary information was examined during this inspection, but is not specifically discussed in this report.
.2 Interim Exit Meetings
- Maintenance Rule Implementation - Periodic Evaluation inspection with Mr. M. Nazar on July 23, 2004.
- Occupational Radiation Safety Radiological Access Controls inspection with Mr. J. Jensen on August 19, 2004.
4OA7 Licensee-Identified Violation
The following violation of very low safety significance was identified by the licensee and was a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation.
Unit 1 and Unit 2 TS 3.7.5.1 required, in part, that the Control Room emergency ventilation system be operable in Modes 1, 2, 3, 4 and during the movement of irradiated fuel assemblies. Condition
- (b) of TS 3.7.5.1 required that with one charcoal adsorber/high efficiency particulate air filter unit inoperable, restore the filter unit to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. Contrary to the above, at 1:37 p.m. on July 19, 2003, with Unit 1 and Unit 2 in Mode 1, the licensee failed to restore the filter units to an operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. This was a violation of TS 3.7.5.1. However, because of the very low safety significance, this violation is being treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy. This issue was discussed in Section 4OA3.1 of this report. The licensee entered this violation into the corrective program as CR 04128075.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- J. Eaton, Maintenance Rule Coordinator
- H. Etheridge, Regulatory Affairs Engineer
- D. Fadel, Vice President Engineering
- M. Finissi, Plant Manager
- J. Gebbie, Engineering Programs Manager
- J. Jensen, Site Vice President
- M. Nazar, Senior Vice President, Chief Nuclear Officer
- R. Serocke, Radiation Protection Manager
- S. Vazquez, System Engineering Manager
- J. Zwolinski, Safety Assurance Director
Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None
Closed
50-315/2004-001-00 LER Failure to Comply with TS 3.7.5.1, Control Room Emergency Ventilation System (Section 4OA3.1)
Discussed
None Attachment