IR 05000424/2008004
| ML083040204 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/29/2008 |
| From: | Scott Shaeffer NRC/RGN-II/DRP/RPB2 |
| To: | Tynan T Southern Nuclear Operating Co |
| References | |
| IR-08-004 | |
| Download: ML083040204 (21) | |
Text
October 29, 2008
SUBJECT:
VOGTLE ELECTRIC GENERATING PLANT - NRC INTEGRATED INSPECTION REPORT 05000424/2008004 AND 05000425/2008004
Dear Mr. Tynan:
On September 30, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Vogtle Electric Generating Plant, Units 1 and 2. The enclosed integrated inspection report documents the inspection findings, which were discussed on October 15, with Mr. D. R. Vineyard and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Based on the results of this inspection, one finding of very low safety significance was identified which was determined to be a violation of regulatory requirements. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section VI.A.1 of the NRC Enforcement Policy because of the very low safety significance of the violation and because it was entered into your corrective action program. If you contest this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Vogtle Electric Generating Plant.
In accordance with the Code of Federal Regulations 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the 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/
Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects
Docket Nos.: 50-424, 50-425 License Nos.: NPF-68 and NPF-81
Enclosures:
Inspection Report 05000424/2008004 and 05000425/2008004 w/Attachment: Supplemental Information
REGION II==
Docket Nos.:
50-424, 50-425
License Nos.:
Report Nos.:
05000424/2008004 and 05000425/2008004
Licensee:
Southern Nuclear Operating Company, Inc.
Facility:
Vogtle Electric Generating Plant, Units 1 and 2
Location:
Waynesboro, GA 30830
Dates:
July 1, 2008 through September 30, 2008
Inspectors:
G. McCoy, Senior Resident Inspector
T. Chandler, Resident Inspector R. Moore, Senior Reactor Inspector (Section 4OA5.2)
C. Even, Reactor Inspector (Section 4OA5.2)
Approved by:
Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000424/2008-004, 05000425/2008-004; 07/01/2008 - 09/30/2008; Vogtle Electric
Generating Plant, Units 1 and 2; Event Followup
The report covered a three-month period of inspection by resident inspectors and two reactor inspectors. One Green non-cited violation was identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Rev 4 dated December 2006.
NRC-Identified and Self-Revealing Findings
- Green.
A self-revealing non-cited violation of TS 5.4, Procedures, was identified when the licensed control room operator failed to perform step 4.3.5.10.b of procedure 3010-1,
Boron Thermal Regeneration System, Rev 57. Failure to properly perform this step resulted in approximately 500 gallons of unborated water being added to the VCT causing an inadvertent positive reactivity addition to the Unit 1 reactor.
The failure to place handswitch 1-HS-7054 in the CLOSED position as required by procedure is a performance deficiency. The minor screening criteria in IMC 0612, Power Reactor Inspection Reports, Appendix B, was not used due to the lack of clear guidance regarding inadvertent reactivity additions. Therefore, in consultation with the NRR program office, regional management determined this finding is more than minor because the licensed operator failed to properly implement procedure 13010-C which resulted an inadvertent positive reactivity addition causing reactor power to briefly increase above 100%. The finding, assessed using the Significance Determination Process, was determined to be of very low safety significance (Green) because the resulting transient did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined this finding was related to the Work Practices component of the Human Performance cross-cutting area in that the licensed operator did not have the procedure in-hand during use as required by procedure 00054-C, Rules for Performing Procedures,
Rev. 19. H.4(b) (Section 4OA3)
Licensee-Identified Violations
None
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at full rated thermal power (RTP) for the entire inspection period.
Unit 2 started the inspection period at essentially full RTP. The unit was shutdown on September 14 for a planned refueling outage.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
==1R01 Adverse Weather Protection
a. Inspection Scope
Impending Adverse Weather Condition.==
On September 4, inspectors reviewed procedures 11889-C, Severe Weather Checklist, and 20054-C, Maintenance Support for the Severe Weather Checklist to verify that the licensee was prepared to implement actions when hurricane Hanna was predicted to affect the Burke County area. The inspectors walked down the outside areas of the plant to identify the presence of loose materials which could affect the operation of plant equipment.
b. Findings
No findings of significance were identified.
==1R04 Equipment Alignment
a. Inspection Scope
Partial System Walkdown.==
The inspectors performed partial walkdowns of the following three systems to verify correct system alignment. The inspectors checked for correct valve and electrical power alignments by comparing positions of valves, switches, and breakers to the documents listed in the Attachment. Additionally, the inspectors reviewed the condition report database to verify that equipment alignment problems were being identified and appropriately resolved.
C Unit 2 auxiliary feedwater (AFW) system prior to testing of the turbine driven AFW pump C
Unit 2 train A high head safety injection pump when the B train was out of service for maintenance C
Unit 2 train A safety injection (SI) pump during a refueling outage
b. Findings
No findings of significance were identified.
==1R05 Fire Protection
a. Inspection Scope
Fire Area Tours.==
The inspectors walked down the following five plant areas to verify the licensee was controlling combustible materials and ignition sources as required by procedures 92015-C, Use, Control, and Storage of Flammable/Combustible Materials, and 92020-C, Control of Ignition Sources. The inspectors assessed the observable condition of fire detection, suppression, and protection systems and reviewed the licensees fire protection Limiting Condition for Operation log and condition report (CR)database to verify that the corrective actions for degraded equipment were identified and appropriately prioritized. The inspectors also reviewed the licensees fire protection program to verify the requirements of Updated Final Safety Analysis Report Section 9.5.1, Fire Protection Program, and Appendix 9A, Fire Hazards Analysis, were met.
Documents reviewed are listed in the Attachment.
C Unit 1 and Unit 2 main control room C
Unit 1 auxiliary building pipe penetration area on level C C
Unit 2 auxiliary building pipe penetration on level B C
Unit 2 auxiliary component cooling water pump room
b. Findings
No findings of significance were identified.
==1R06 Flood Protection Measures
a. Inspection Scope
Internal Flood Review.==
The inspectors walked down the following area which contained risk-significant structures, systems and components below flood level to verify flood barriers were in place. Motor controllers and terminal boxes that could become potentially submerged were inspected to ensure that the sealing gasket material was intact and undamaged. The inspectors reviewed selected licensee alarm response procedures to verify alarm setpoints and setpoints for sump pump operation were consistent with the UFSAR, the setpoint index, and Technical Specifications (TSs).
C Unit 2 AFW building
b. Findings
No findings of significance were identified.
==1R11 Licensed Operator Requalification
a. Inspection Scope
Resident Quarterly Observation.==
The inspectors observed operator performance on July 16, during licensed operator simulator training described on simulator exercise guide Dynamic Simulator Scenarios V-RQ-SE-08405. The simulator scenarios covered operator actions resulting from steam generator tube rupture, a pressurizer steam space leak, and an automatic reactor trip failure. Documents reviewed are listed in the
. The inspectors specifically assessed the following areas:
C Correct use of the abnormal and emergency operating procedures C
Ability to identify and implement appropriate actions in accordance with the requirements of the technical specifications C
Clarity and formality of communications in accordance with procedure 1000-C, Conduct of Operations C
Proper control board manipulations including critical operator actions C
Quality of supervisory command and control C
Effectiveness of the post-evaluation critique
b. Findings
No findings of significance were identified.
==1R12 Maintenance Effectiveness
a. Inspection Scope
==
The inspectors reviewed the following two condition reports to evaluate the effectiveness of the licensees handling of equipment performance problems and to verify the licensees maintenance efforts met the requirements of 10 CFR 50.65 (the Maintenance Rule) and licensee procedure 50028-C, Engineering Maintenance Rule Implementation.
The inspector also reviewed one safety-significant system to verify that the licensees maintenance efforts met the requirements of 10 CFR 50.65 (the Maintenance Rule) and licensee procedure 50028-C, Engineering Maintenance Rule Implementation. The reviews included adequacy of the licensees failure characterization, establishment of performance criteria or 50.65(a)(1) performance goals, and adequacy of corrective actions. Other documents reviewed during this inspection included control room logs, system health reports, the maintenance rule database, and maintenance work orders.
Also, the inspectors interviewed system engineers and the maintenance rule coordinator to assess the accuracy of identified performance deficiencies and extent of condition.
C CR 2008105578, Containment cooling fan number 1 failed to start in slow speed.
C CR 2008106489, Unit 2 train B NSCW fan number 4 tripped.
b. Findings
No findings of significance were identified.
==1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
==
The inspectors reviewed the following four work activities to verify plant risk was properly assessed by the licensee prior to conducting the activities. The inspectors reviewed risk assessments and risk management controls implemented for these activities to verify they were completed in accordance with procedure 00354-C, Maintenance Scheduling, and 10 CFR 50.65(a)(4). The inspectors also reviewed the CR database to verify that maintenance risk assessment problems were being identified at the appropriate level, entered into the corrective action program, and appropriately resolved.
C July 25, planned maintenance outages for the Unit 2 train A containment spray (CS)pump and the Unit 2 Train A SI pump C
July 28, Unit 2 train B coolant charging pump was out of service for planned maintenance C
August 4, Unit 2 train A CS pump was out of service for corrective maintenance C
September 6, performance of multiple surveillances in preparation for Unit 2 refueling outage
b. Findings
No findings of significance were identified.
==1R15 Operability Evaluations
a. Inspection Scope
==
The inspectors reviewed the following five evaluations to verify they met the requirements of procedure NMP-GM-002, Corrective Action Program, and NMP-GM-002-001, Corrective Action Program Instructions. The scope of this inspection included a review of the technical adequacy of the evaluations, the adequacy of compensatory measures, and the impact on continued plant operation.
C CR 2008107647, Unit 2 train A sequencer power supply failure.
C CR 2008107492, Fuel assembly in the spent fuel pool lifted by the hold down spring C
CR 2008109254, Defective security diesel generator breaker C
CR 2008109286, Breaches in the Unit 2 piping penetration negative pressure boundary C
CR 2008109859, ultra-low sulfur diesel fuel was added to the 2B EDG fuel oil storage tank.
b. Findings
No findings of significance were identified.
==1R19 Post-Maintenance Testing
a. Inspection Scope
==
The inspectors either observed post-maintenance testing or reviewed the test results for the following four maintenance activities to verify that the testing met the requirements of procedure 29401-C, Work Order Functional Tests, for ensuring equipment operability and functional capability was restored. The inspectors also reviewed the test procedures to verify the acceptance criteria were sufficient to meet the (TS) operability requirements.
C Work order (WO) 10813359, Replacement of failed Unit 1 Train B reactor trip breaker C
WOs 20601148, 20802972, and 20802974, Unit 2 train A CS pump Outage C
WO 20715268, Clean/inspect/test/lubricate the Unit 2 train B NSCW pump #4 discharge valve (2HV11613)
C WO 20810774, Unit 2 train B CS pump mechanical seal replacement
b. Findings
No findings of significance were identified.
==1R20 Refueling and Other Outage Activities
a. Inspection Scope
==
The inspectors performed the inspection activities described below for the Unit 2 refueling outage that began on September 14. The inspectors confirmed that, when the licensee removed equipment from service, the licensee maintained defense-in-depth commensurate with the outage risk control plan for key safety functions and applicable technical specifications and that configuration changes due to emergent work and unexpected conditions were controlled in accordance with the outage risk control plan.
Documents reviewed are listed in the Attachment. Inspection activities included:
C Prior to the outage, the resident inspectors reviewed the licensees integrated risk control plan to verify that activities, systems, and/or components which could cause unexpected reactivity changes were identified in the outage risk plan.
C Observed portions of the plant shutdown and cooldown to verify that the technical specification cooldown restrictions were followed.
C Reviewed reactor coolant system pressure, level, and temperature instruments to verify that the instruments provided accurate indication and that allowances were made for instrumentation errors C
Verified that outage work did not impact the operation of the spent fuel cooling system during and after core load.
C Reviewed the status and configuration of electrical systems to verify that those systems met technical specification requirements and the licensees outage risk control plan C
Observed decay heat removal parameters to verify that the system was properly functioning and providing cooling to the core C
Reviewed system alignments to verify that the flow paths, configurations and alternative means for inventory addition were consistent with the outage risk plan C
Reviewed selected control room operations to verify that the licensee was controlling reactivity in accordance with the technical specifications C
Observed the licensees control of containment penetrations to verify that the requirements of the technical specifications were met C
Reviewed the licensees plans for changing plant configuration to verify that technical specifications, license conditions, and other requirements, commitments, and administrative procedure prerequisites were met prior to changing plant configuration
b. Findings
No findings of significance were identified.
==1R22 Surveillance Testing
a. Inspection Scope
==
The inspectors reviewed the following four surveillance test procedures and either observed the testing or reviewed test results to verify that testing was conducted in accordance with the procedures and that the acceptance criteria adequately demonstrated that the equipment was operable. Additionally, the inspectors reviewed the CR database to verify that the licensee had adequately identified and implemented appropriate corrective actions for surveillance test problems.
Surveillance Tests C
14980A-2, Unit 2 train A EDG monthly operability test
In-Service Tests (IST)
C 14545-2, Unit 2 train A MDAFW pump monthly operability test C
14545-2, Unit 2 train B MDAFW pump monthly operability test C
14802-1, Unit 1 B train NSCW pump #6 IST and response time test
b. Findings
No findings of significance were identified.
1EP6 Drill Evaluation
a. Inspection Scope
The inspectors reviewed the facility activation exercise guide and observed the following emergency response activity to verify the licensee was properly classifying emergency events, making the required notifications, and making appropriate protective action recommendations in accordance with procedures 91001-C, Emergency Classifications, and 91305-C, Protective Action Guidelines.
C On August 6 the licensee conducted an emergency preparedness drill involving an uncontrolled release of activity from a waste decay tank, followed by a loss of all offsite power and a loss of all onsite AC power. The technical support center was activated and the site participated in the exercise.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
a. Inspection Scope
The inspectors sampled licensee submittals for the listed PIs during the period from July 1, 2007 - June 30, 2008, for Unit 1 and Unit 2. The inspectors verified the licensees basis in reporting each data element using the PI definitions and guidance contained in procedures 00163-C, NRC Performance Indicator and Monthly Operating Report Preparation and Submittal, and Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Indicator Guideline.
Mitigating Systems Cornerstone C
Mitigating Systems Performance Index, High Pressure Injection System C
Mitigating Systems Performance Index, Residual Heat Removal System C
Mitigating Systems Performance Index, Heat Removal System
The inspectors reviewed Unit 1 and Unit 2 operator log entries, the Vogtle MSPI basis document, the monthly operating reports and monthly PI summary reports to verify that the licensee had accurately submitted the PI data.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
.1 Daily Condition Report Review.
As required by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program. This review was accomplished by either attending daily screening meetings that briefly discussed major CRs, or accessing the licensees computerized corrective action database and reviewing each CR that was initiated.
.2 Focused Review
a. Inspection Scope
The inspectors performed a detailed review of the following CR to verify the full extent of the issue was identified, an appropriate evaluation was performed, and appropriate corrective actions were specified and prioritized. The inspectors evaluated the CR against the licensees corrective action program as delineated in licensee procedure NMP-GM-002, Corrective Action Program, and 10 CFR 50, Appendix B.
C CR 2008106941, During a surveillance test the inboard seal of the Unit 2 train B CS pump overheated.
b. Findings
No findings of significance were identified.
4OA3 Event Followup
.1 Inadvertent Positive Reactivity Addition
a. Inspection Scope
Resident inspectors reviewed the circumstances surrounding an unintended dilution of the reactor coolant system as described in condition report 2008108672, Inadvertent dilution while performing procedure for chemistry control of the Boron Thermal Regeneration System (BTRS) chilled water system. The inspectors interviewed operators on-shift at the time of the event, reviewed the condition report and subsequent disposition, attended the associated human performance board, reviewed the applicable procedures, and evaluated the plant response using plots of reactor power and Tave during the event.
b. Findings
Introduction.
A Green self-revealing NCV of TS 5.4, Procedures, was identified when the licensed control room operator failed to perform step 4.3.5.10.b of procedure 3010-1, Boron Thermal Regeneration System, Rev 57. Failure to properly perform this step resulted in approximately 500 gallons of unborated water being added to the VCT, causing an inadvertent positive reactivity addition to the Unit 1 reactor.
Description.
On August 19, the licensee was conducting a flush of the BTRS chilled water piping using procedure 13010-1. This procedure was designated as Continuous Use which required operators to have the procedure in-hand during use. The flush was being directed by a field operator because the majority of the operator actions were performed outside the main control room. However, a few steps were performed in the main control room by the licensed operator who was designated as Operator at the Controls. Due to the limited number of steps to be performed, the licensed operator decided that it was not necessary to have the procedure in-hand. Step 4.3.5.10.b directed the CVCS BTRS inlet isolation valve handswitch, 1-HS-7054, be placed in the CLOSED position. When directed by the field operator to perform step 4.3.5.10.b, the licensed operator verified that valve 1-HV-7054 was closed, but did not place handswitch 1-HS-7054 in the CLOSED position as required by procedure 13010-1. Subsequently, when the licensed operator placed the CVCS BTRS selector switch in the DILUTE position, valve 1-HV-7054 automatically opened. As a result, approximately 500 gallons of unborated water from the BTRS was added to the VCT. The addition of unborated water to the VCT resulted in an inadvertent positive reactivity addition to the Unit 1 reactor which caused power to increase above 100% RTP. The operators became aware of the reactivity addition when control rods began to automatically insert. The control room operators responded to the event by manually inserting control rods and reducing main generator load to reduce reactor power below 100%. The maximum 1-minute average power achieved was 100.73%.
Analysis.
The failure to place handswitch 1-HS-7054 in the CLOSED position as required by procedure is a performance deficiency. The minor screening criteria in IMC 0612, Power Reactor Inspection Reports, Appendix B, was not used due to the lack of clear guidance regarding inadvertent reactivity additions. Therefore, in consultation with the NRR program office, regional management determined this finding is more than minor because the licensed operator failed to properly implement procedure 13010-C which resulted in an inadvertent positive reactivity addition causing reactor power to briefly increase above 100%. This finding, assessed using the SDP, was determined to be of very low safety significance (Green) because the resulting transient did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined this finding was related to the Work Practices component of the Human Performance cross-cutting area in that the licensed operator did not have the procedure in-hand during use as required by procedure 00054-C, Rules for Performing Procedures, Rev. 19. H.4(b)
Enforcement.
Technical Specification 5.4.1.a. requires that written procedures as specified in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, be established, implemented, and maintained. Regulatory Guide 1.33 states that instructions for energizing, filling, venting, draining, startup, shutdown, and changing modes of operation should be prepared for the CVCS, which contains the BTRS.
Contrary to the above, on August 19, 2008, a licensed reactor operator failed to implement procedure 13010-1 when he improperly performed step 4.3.5.10.b. Because the finding was of very low safety significance and was entered into the licensees corrective action program as CR 2008108672, this violation is being treated as an NCV, consistent with section VI.A.1 of the NRC Enforcement Policy: NCV 05000424/2008004-01, Human Performance Error Results in Unplanned Reactivity Addition.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel and Activities
a. Inspection Scope
During the inspection period, the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours.
b. Findings
No findings of significance were identified.
.2 (Discussed)
Temporary Instruction (TI) 2515/176, Emergency Diesel Generator Technical Specification Surveillance Requirements Regarding Endurance and Margin Testing
a. Inspection Scope
The objective of this TI was to gather information to assess the adequacy of nuclear power plant EDG endurance and margin testing as prescribed by plant-specific TSs.
The inspector interfaced with the appropriate station staff to obtain the information specified in Attachment 1 of the TI Worksheet. The TI applies to all operating nuclear power reactor licensees that use EDGs as the onsite standby power supply. The inspector verified the accuracy of the information by review of TS, EDG Design Basis Event (DBE) loading calculations, EDG endurance run test procedures, test data from the last three endurance tests performed on each EDG, EDG ratings, and EDG operating history. The information gathered will be forwarded to Nuclear Reactor Regulation/Division of Engineering/Electrical Engineering Branch (NRR/DE/EEEB) for further review to assess the adequacy and consistency of EDG testing at nuclear stations.
b. Findings and Observations
The TI is presently scheduled to be open until completion of the NRR/DE/EEEB review.
.3 Review of Institute of Nuclear Power Operations (INPO) Evaluation Report
The inspectors reviewed the results of an INPO evaluation of licensee performance conducted during January and February 2008. The report did not identify any significant licensee performance issues that had not been previously addressed and/or reviewed by the NRC.
4OA6 Meetings, Including Exit
Exit Meeting
On October 15, the resident inspectors presented the inspection results to Mr. D. R.
Vineyard and other members of his staff, who acknowledged the findings. The inspectors confirmed that proprietary information was not provided or examined during the inspection.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- R. Brown, Training and Emergency Preparedness Manager
- C. Buck, Chemistry Manager
- W. Copeland, Performance Analysis Supervisor
- R. Dedrickson, Plant Manager
- K. Dyar, Security Manager
- I. Kochery, Health Physics Manager
- J. Robinson, Work Control Superintendent
- T. Tynan, Site Vice-President
- D. Vineyard, Operations Manager
- J. Williams, Site Support Manager
- T. Youngblood, Site Engineering Manager
NRC personnel
- S. Shaeffer, Chief, Region II Reactor Projects Branch 2
LIST OF ITEMS
OPENED AND CLOSED
Opened and Closed
- 05000424/2008004-01 NCV Human Performance Error Results in Unplanned Reactivity Addition
Discussed
2515/176 TI Emergency Diesel Generator Technical Specification Surveillance Requirements Regarding Endurance and Margin Testing