IR 05000338/2011002
| ML111220520 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/02/2011 |
| From: | Gerald Mccoy NRC/RGN-II/DRP/RPB5 |
| To: | Heacock D Virginia Electric & Power Co (VEPCO) |
| References | |
| IR-11-002 | |
| Download: ML111220520 (25) | |
Text
May 2, 2011
SUBJECT:
NORTH ANNA POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000338/2011002 AND 05000339/2011002
Dear Mr. Heacock:
On March 31, 2011, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your North Anna Power Station Units 1 and 2. The enclosed integrated inspection report documents the inspection findings which were discussed on April 21, 2011, with Mr. Fred Mladen, on April 25, 2011, and May 2, 2011, with Mr. Don Taylor, and other members of your staff.
The inspection examined activities conducted under your licenses as they related to safety and compliance with the Commissions rules and regulations and with the conditions of your licenses. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
This report documents two NRC-identified and one self-revealing findings of very low safety significance (Green) which were determined to be violations of NRC requirements. However, because of the very low safety significance of these issues and because they were entered into your corrective action program, the NRC is treating these as non-cited violations (NCV)
consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you wish to contest these non-cited violations, 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 20555-001; 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 North Anna Power Station.
Additionally, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the North Anna Power Station. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, 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/
Gerald J. McCoy, Chief
Reactor Projects Branch 5
Division of Reactor Projects
Docket Nos. 50-338, 50-339 License Nos. NPF-4, NPF-7
Enclosure:
Inspection Report 05000338/2011002 and 05000339/2011002
w/ Attachment: Supplemental Information
REGION II==
Docket Nos.:
50-338, 50-339
License Nos.:
Report No.:
05000338/2011002, 05000339/2011002
Licensee:
Virginia Electric and Power Company (VEPCO)
Facility:
North Anna Power Station, Units 1 & 2
Location:
1022 Haley Drive Mineral, Virginia 23117
Dates:
January 1, 2011 through March 31, 2011
Inspectors:
J. Reece, Senior Resident Inspector
R. Clagg, Resident Inspector
C. Sanders, Acting Resident Inspector
Approved by:
Gerald J. McCoy, Chief Reactor Projects Branch 5 Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000338/2011-002, 05000339/2011-002; 01/01/2011 - 03/31/2011; North Anna Power
Station, Units 1 and 2: Operability Evaluations, Post Maintenance Testing, Identification and Resolution of Problems and Event Followup.
The report covered a 3 month period of inspection by resident inspectors. Three findings were identified and were determined to be non-cited violations (NCVs). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). The cross-cutting aspect was determined using IMC 0310, Components Within the Cross Cutting Areas. Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process Revision 4, dated December 2006.
NRC Identified and Self-Revealing Findings
Cornerstone: Initiating Events
- Green.
A non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," was identified by the inspectors for the licensees failure to determine the cause of a significant condition adverse to quality (SCAQ) involving an automatic reactor trip following a lightning strike on the Unit 2 containment building. This resulted in the Unit 2 automatic reactor trip on June 16, 2010, because of the insufficient corrective action to preclude repetition. The Licensee entered this issue into the Corrective Action Program as CR 384967.
The inspectors determined that the failure to determine the cause of a SCAQ was a performance deficiency (PD). The inspectors reviewed IMC 0612, Appendix B and determined the PD was more than minor because, if left uncorrected, it has the potential to lead to a more significant safety concern in that failing to identify the cause of SCAQs and thus failing to take corrective action to preclude repetition could result in additional initiating events or impacts on mitigating systems. In addition, the inspectors determined that it adversely impacted the Initiating Events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, specifically the attribute of Protection Against External Factors in that the removal of the Overtemperature Delta T lag function removed protection from lightning strikes on the reactor protection system. The inspectors reviewed IMC 0609, Attachment 4 and determined that the finding was of very low safety significance, or Green, because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The cause of this finding involved the cross-cutting area of problem identification and resolution, the component of operating experience, and the aspect of evaluation of identified problems, P.1(c) because the licensee failed to thoroughly evaluate the cause of the 2005 reactor trip and conduct effectiveness reviews of corrective actions to ensure the problems are resolved. (Section 4OA3.1)
Cornerstone: Mitigating Systems
- Green.
A non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions,
Procedures and Drawings, was identified by the NRC for the failure to accomplish the installation of Unit 2 low head safety injection (LHSI) piping and supports in accordance with prescribed drawings which resulted in no contact between piping and two different pipe supports and caused an operable but degraded and nonconforming condition. The licensee entered this problem into their corrective action program as condition reports 413315 and 418989.
A performance deficiency was identified by the NRC for the failure to adequately install Unit 2 LHSI pipe supports in accordance with prescribed drawings. This PD had a credible impact on safety due to the loss of design basis margin resulting in a reasonable doubt regarding reliability and capability during a seismic event. The PD was more than minor because it impacted the mitigating systems cornerstone objective to ensure the reliability and capability of systems which respond to initiating events and the related attribute of equipment performance because the reliability of the support configurations had been impacted by the reduction in design margin. In accordance with NRC IMC 0609, Significant Determination Process, the inspectors performed a Phase 1 analysis and determined the finding was of very low safety significance or Green due to a design deficiency confirmed not to result in a loss of operability or functionality. The finding had no cross-cutting aspects due to its legacy nature. (Section 1R15)
- Green.
A Green, non-cited violation of 10 CFR 50, Appendix B, Criterion V,
"Instructions, Procedures, and Drawings," was identified by the NRC for failure to adequately prescribe the correct program instructions to ensure safety-related instrument and control (I&C) preventative maintenance (PMs) received the appropriate post maintenance testing (PMT). The licensee entered this problem into their corrective action program as condition report 417730.
A performance deficiency was identified by the NRC for the failure to adequately prescribe programmatic PMT instructions to ensure safety-related I&C PMs had proper PMT. The inspectors reviewed Inspection Manual Chapter (IMC) 0612,
Appendix BProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0612,</br></br>Appendix B" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., and determined the finding was more than minor because if left uncorrected it would have the potential to result in a more significant safety event. In accordance with IMC 0609, Significant Determination Process, the inspectors performed a Phase 1 analysis and determined that the finding was of very low significance because the finding was not a design deficiency, did not represent a loss of safety function and did not screen as potentially risk significant due to a seismic, flooding or severe weather initiating event. This finding involved the cross-cutting area of human performance, the component of the resources, and the aspect of complete documentation, H.2(c), because the licensee failed to adequately prescribe programmatic PMT instructions to ensure safety-related I&C PMs had proper PMT.
(Section 1R19)
Licensee Identified Violations
None
REPORT DETAILS
Summary of Plant Status
Unit 1 began the period at full Rated Thermal Power (RTP) and operated at or near full RTP for the report period with the following exceptions. On January 6, 2011, the unit reduced power to 85% for a main turbine governor electrohydraulic valve replacement. On February 9, 2011, the unit reduced power to 88% for main turbine governor control wiring repairs.
Unit 2 began the period at full RTP and operated at or near full RTP for the report period with the following exception. On January 7, 2011, the unit reduced power to 85% for main turbine condenser leak repairs.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
==1R01 Adverse Weather Protection
a. Inspection Scope
==
On February 28, 2011, a tornado watch was issued for Louisa County and the inspectors performed a reactive weather related inspection. The inspectors reviewed licensee adverse weather response procedure, 0-AP-41, Severe Weather, Revision 52, and related site preparations including work activities that could impact the overall maintenance risk assessments.
b. Findings
No findings were identified.
==1R04 Equipment Alignment
a. Inspection Scope
==
The inspectors conducted four equipment alignment partial walkdowns to evaluate the operability of selected redundant trains or backup systems, listed below, with the other train or system inoperable or out of service. The inspectors reviewed the functional systems descriptions, Updated Final Safety Analysis Report (UFSAR), system operating procedures, and Technical Specifications (TS) to determine correct system lineups for the current plant conditions. The inspectors performed walkdowns of the systems to verify that critical components were properly aligned and to identify any discrepancies which could affect operability of the redundant train or backup system.
- Unit 2 A Motor Driven (MD) Auxiliary Feedwater (AFW) and Turbine Driven AFW during 5 year preventive maintenance on 'B MDAFW
- Unit 2 2J Emergency Diesel Generator (EDG) during emergent work on 2H EDG
- Unit 2 A Quench Spray Pump (QSP) walkdown during planned maintenance on the B QSP
b. Findings
No findings were identified
==1R05 Fire Protection
a. Inspection Scope
==
The inspectors conducted tours of the five areas listed below that are important to reactor safety to verify the licensees implementation of fire protection requirements as described in fleet procedures CM-AA-FPA-100, Revision 22, Fire Protection/Appendix R (Fire Safe Shutdown) Program, CM-AA-FPA-101, Control of Combustible and Flammable Materials, Revision 3, and CM-AA-FPA-102, Fire Protection and Fire Safe Shutdown Review and Preparation Process and Design Change Process, Revision 1.
The inspectors evaluated, as appropriate, conditions related to:
- (1) licensee control of transient combustibles and ignition sources;
- (2) the material condition, operational status, and operational lineup of fire protection systems, equipment, and features; and
- (3) the fire barriers used to prevent fire damage or fire propagation.
- Emergency Diesel Generator 1H Unit 1 (fire zone 9A-1a / EDG-1H) and Emergency Diesel Generator 2H Unit 2 (fire zone 9A-2a / EDG 2H)
- Emergency Diesel Generator 1J Unit 1 (fire zone 9B-1a / EDG-1J) and Emergency Diesel Generator 2J Unit 2 (fire zone 9B-2a / EDG-2J)
- Cable Vault and Tunnel Unit 1 (includes Control Rod Drive Room and Z-27-1)(fire zone 3-1a / CV&T-1)
- Normal Switchgear Room Unit 1 and Unit 2 (fire zones 5-1 / NSR-1 and 5-2 / NSR-2)
- Battery Rooms 1-1, 1-II, 1-III, and 1-IV Unit 1 (fire zones 7A-1 / BR1-I, 7B-1 / BR 1-II, 7C-1 / BR1-III, and 7D-1 / BR1-IV) and Battery Rooms 2-I, 2-II, 2-III and 2-IV Unit 2 (fire zones 7A-2 / BR2-I, 7B-2 / BR 2-II, 7C-2 / BR2-III, and 7D-2 / BR2-IV)
b. Findings
No findings were identified.
==1R06 Flood Protection Measures
a. Inspection Scope
==
The inspectors performed an annual review of cables located in underground bunkers/manholes. The inspectors evaluated, as appropriate, the cable vaults listed below for the following:
- (1) verified by direct observation that the cables were not submerged in water;
- (2) verified by direct observation that cables and/or splices appeared intact;
- (3) verified that drainage or an appropriate dewatering device (sump pump) was in operation; and
- (4) verified that level alarm circuits were set appropriately to ensure that the cables would not be submerged.
- 01-BLD-MBAR-5MH03
- 01-BLD-MBAR-5MH04
- 01-BLD-MBAR-50MH-3
- 01-BLD-MBAR-50MH-4
b. Findings
No findings were identified.
==1R11 Licensed Operator Requalification Program
==
Operating Experience Smart Sample (OpESS) FY 2010-02 Sample Selections for Reviewing Licensed Operator Examinations and Training Conducted on the Plant-Reference Simulator
a. Inspection Scope
The inspectors observed an operator requalification complex situation simulator scenario which involved a loss of main condenser vacuum, failure of the control rods to move in automatic, secondary system transients induced by multiple failures of heater drain pumps, a failure of the reactor to trip on a manual attempt, control rod failure, shaft failure of a motor driven AFW pump, loss of an emergency 4160V bus and respective EDG, and a faulted steam generator.
The inspectors observed crew performance in terms of communications; ability to prioritize failures in order to take timely and proper actions; prioritizing, interpreting, and verifying alarms; correct use and implementation of procedures, including the alarm response procedures; timely control board operation and manipulation, including high-risk operator actions; and oversight and direction provided by the shift supervisor, including the ability to identify and implement appropriate TS actions. The inspectors observed the post training critique to determine that weaknesses or improvement areas revealed by the training were captured by the instructor and reviewed with the operators.
b. Findings
No findings were identified.
==1R12 Maintenance Effectiveness
a. Inspection Scope
==
For the three equipment issues listed below, the inspectors evaluated the effectiveness of the respective licensee's preventive and corrective maintenance. The inspectors performed walkdowns of the accessible portions of the systems, performed in-office reviews of procedures and evaluations, and held discussions with licensee staff. The inspectors compared the licensees actions with the requirements of the Maintenance Rule (10 CFR 50.65), and licensee procedure ER-AA-MRL-10, Maintenance Rule Program, Revision 4.
- MRE 013094, Annnunciator J-C4 received for 2-RS-P-2A seal head tank low level
- MRE 013005, MRE to engineering to address unanalyzed scaffolding in 1-CH-P-1A
b. Findings
The enforcement aspects for scaffolding in the 1-CH-P-1A room are discussed in section 4OA2.2 of this report.
==1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
==
The inspectors evaluated, as appropriate, the four activities listed below for the following:
- (1) effectiveness of the risk assessments performed before maintenance activities were conducted;
- (2) management of risk;
- (3) upon identification of an unforeseen situation, necessary steps were taken to plan and control the resulting emergent work activities; and
- (4) maintenance risk assessments and emergent work problems were adequately identified and resolved. The inspectors verified that the licensee was in compliance with the requirements of 10 CFR 50.65 (a)(4) and the data output from the licensees safety monitor associated with the risk profile of Units 1 and 2.
- Emergent down power on Unit 2 for main turbine condenser waterbox tube leaks
- Emergent work on Unit 2 '2H' EDG to replace breaker for pre-lube pump
- Emergent work on 2-RS-P-2A due to low seal head tank level
- Entry into 0-AP-41, Severe Weather Condition, Revision 52, on February 28, 2011 due to issuance of a tornado watch
b. Findings
No findings were identified.
==1R15 Operability Evaluations
a. Inspection Scope
==
The inspectors reviewed five operability evaluations, listed below, affecting risk-significant mitigating systems, to assess, as appropriate:
- (1) the technical adequacy of the evaluations;
- (2) whether continued system operability was warranted;
- (3) whether other existing degraded conditions were considered as compensating measures; (4)whether the compensatory measures, if involved, were in place, would work as intended, and were appropriately controlled; and
- (5) where continued operability was considered unjustified, the impact on TS Limiting Conditions for Operation and the risk significance in accordance with the Significant Determination Process (SDP). The inspectors review included a verification that operability determinations (OD) were made as specified by procedure OP-AA-102, Operability Determination, Revision 6.
- CR410129, Annuciator J-C4 received for 2-RS-P-2A tank low level"
- CR412153, review of OD 000402/403, Evaluate Operability of Masoneilan Valve Actuator Diaphragms
- CR413315, Unit 2 SI pipe support not in contact with pipe
- CR413453, 1H EDG air start pilot line is contacting the exhaust system
- CR418989, Unit 2 SI pipe support 2-SI-R-5 not in contact with pipe
b. Findings
Inadequate Installation of Unit 2 Low Head Safety Injection Piping and Related Supports
Introduction:
A Green, NCV of 10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings," was identified by the NRC for failure to accomplish the installation of Unit 2 low head safety injection (LHSI) piping and supports in accordance with prescribed drawings which resulted in no contact between piping and two different pipe supports and caused an operable but degraded and nonconforming condition.
Description:
On February 9, 2011, the inspectors identified Unit 2 LHSI pipe, 8"-SI-449-153A-Q2, not in contact with support, 2-SI-R-214, as required by drawing, 12050-PSSK-107K.11. The licensee initiated CR413315 and associated operability determination (OD) 000404. On March 23, 2011, the inspectors identified Unit 2 LHSI pipe, 12-SI-405-153A-Q3, not in contact with support, 2-SI-R-5, as required by drawing, 12050-PSSK-104C.08. The licensee initiated CR418989 and associated OD000409. The inspectors reviewed both of the ODs which concluded each configuration resulted in an operable but not fully qualified condition. Additionally, each OD concluded that based on the dead weight of the pipe not supported by the deficiencies and consequently distributed to adjacent supports, the design margin for the adjacent supports had been reduced which required corrective action to return each pipe/support configuration to the design requirements.
The inspectors determined that both pipe/support configurations caused a degraded condition based on the reduction of design basis margin and a nonconforming condition based on noncompliance with the respective design drawings and related stress analysis.
Analysis:
The inspectors determined that the failure to install the aforementioned Unit 2 LHSI piping and supports in accordance with design drawings was a performance deficiency (PD). This PD had a credible impact on safety due to the loss of design basis margin resulting in a reasonable doubt regarding reliability and capability during a seismic event. The inspectors reviewed Inspection Manual Chapter (IMC) 0612, Appendix B, and determined the finding was more than minor because impacted the mitigating systems cornerstone objective to ensure the reliability and capability of systems which respond to initiating events and the related attribute of equipment performance because the reliability of the support configurations had been impacted by the reduction in design margin. In accordance with NRC IMC 0609, Significant Determination Process, the inspectors performed a Phase 1 analysis and determined the finding was of very low safety significance or Green due to a design deficiency confirmed not to result in a loss of operability or functionality. The finding had no cross-cutting aspects due to its legacy nature.
Enforcement:
10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," states in part that activities affecting quality shall be accomplished by documented instructions and procedures. Contrary to the above, on February 9 and March 23, 2011, inspectors identified that the licensee failed to accomplish the correct installation of Unit 2 LHSI piping and supports 2-SI-R-5 and 2-SI-R-214 in accordance with documented, design drawings which caused a degraded and nonconforming condition. Because the finding is of very low safety significance and it was entered into the licensees corrective action program (CAP) as CR413315 and CR418989, this violation is being treated as a Green NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000339/2011002-01, Inadequate Installation of Unit 2 Low Head Safety Injection Piping and Related Supports.
==1R19 Post Maintenance Testing
a. Inspection Scope
==
The inspectors reviewed six post maintenance test procedures and/or test activities for selected risk-significant mitigating systems listed below, to assess whether:
- (1) the effect of testing on the plant had been adequately addressed by control room and/or engineering personnel;
- (2) testing was adequate for the maintenance performed; (3)acceptance criteria were clear and adequately demonstrated operational readiness consistent with design and licensing basis documents;
- (4) test instrumentation had current calibrations, range, and accuracy consistent with the application;
- (5) tests were performed as written with applicable prerequisites satisfied;
- (6) jumpers installed or leads lifted were properly controlled;
- (7) test equipment was removed following testing; and
- (8) equipment was returned to the status required to perform in accordance with VPAP-2003, Post Maintenance Testing Program, Revision 13.
- WO 59102235793, Reactivity management adjust valve stroke for 1-CH-FCV-1114A
- WO 59101657315, Calibrate 02-EG-PS-703J
- WO 59102275772, Inspect and adjust MOP in accordance with 0-ECM-0708-03 on 2J EDG
- WO 59102276225, Replace EDG breaker control switch
- WO 59102257415, Adjust closed limit to electrically seat 1-SW-MOV-110B
- WO 59102226732, Replace 1A charging pump auxiliary oil pump
b. Findings
Introduction:
A Green, NCV of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified by the NRC for failure to adequately prescribe the correct program instructions to ensure safety-related instrument and control (I&C) preventative maintenance (PMs) received the appropriate post maintenance testing (PMT).
Description:
On March 4, 2011, during a review of work order (WO) 59101657315 for PM involving calibration of 02-EG-PS-703J, 2J EDG fuel oil low pressure switch, the inspectors identified that no PMT was performed. The inspectors interviewed engineering personnel and reviewed related PMT program documentation including NASES-4.17, Controlling Procedure for the Post Maintenance Testing Program, Revision 0, of which Attachment 2 contained the following, NOTE: Only I&C corrective WOs require PMT TDs, I&C PMs are excluded at this time. The licensee indicated this note was used to during the WO planning process to not require a PMT. The inspectors questioned the validity of this note because the instrument tubing for the pressure switch had been disconnected for the calibration. The licensee personnel interviewed agreed that a PMT requiring an external leakage check should have been stipulated and that the program procedure would have to be revised.
The inspectors noted that the EDG pressure switch was located within a control cabinet, which was not part of the normal leakage inspection areas by operations, and that fuel oil leakage within the control cabinet can contribute to a more significant safety event involving a fire. The inspectors concluded that the licensee program procedure, NASES-4.17, failed to adequately prescribe PMT instructions to ensure safety-related I&C PMs had proper PMT. On March 15, 2011, the licensee entered this issue into their CAP as CR417730.
Analysis:
The inspectors determined that the failure to adequately prescribe programmatic PMT instructions to ensure safety-related I&C PMs had proper PMT was a performance deficiency (PD). The inspectors reviewed IMC 0612, Appendix B, and determined the finding was more than minor because if left uncorrected it would have the potential to result in a more significant safety event. Since the program instructions did not require adequate PMT for safety-related I&C PM activities, then for the case noted above involving the EDG fuel oil pressure switch external leakage would have to potential to lead to an EDG fire scenario. In accordance with IMC 0609, Significant Determination Process, the inspectors performed a Phase 1 analysis and determined that the finding was of very low significance because the finding was not a design deficiency, did not represent a loss of safety function and did not screen as potentially risk significant due to a seismic, flooding or severe weather initiating event. This finding involved the cross-cutting area of human performance, the component of the resources, and the aspect of complete documentation, H.2(c), because the licensee failed to maintain accurate and up to date procedures and work packages for PMTs following safety-related I&C PM.
Enforcement:
10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," states in part that activities affecting quality shall be prescribed by documented instructions and procedures. Contrary to the above, on March 15, 2011, the licensees procedure, NASES-4.17, failed to adequately prescribe the correct instructions to ensure safety-related I&C PMs received the appropriate PMT. Because the finding is of very low safety significance and it was entered into the licensees CAP as CR417730, this violation is being treated as a Green NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000338, 339/2011002-02, Inadequate Post Maintenance Test Program Instructions for Safety-Related Instrument and Control Preventative Maintenance.
==1R22 Surveillance Testing
a. Inspection Scope
==
For the five surveillance tests listed below, the inspectors examined the test procedures, witnessed testing, or reviewed test records and data packages, to determine whether the scope of testing adequately demonstrated that the affected equipment was functional and operable, and that the surveillance requirements of TS were met. The inspectors also determined whether the testing effectively demonstrated that the systems or components were operationally ready and capable of performing their intended safety functions.
In-Service Test:
- 1-PT-75.2A, Service Water Pump (1-SW-P-1A) Quarterly Test, Revision 51
- 1-PT-63.1A, Quench Spray System - A Subsystem, Revision 36
- 1-PT-14.1, Charging Pump 1-CH-P-1A, Revision 48
RCS Leakage:
- 2-PT-52.2A, Reactor Coolant System Leak rate (Computer Calculation), Revision Other Surveillance Tests:
- 1-PT-36.1A, Train A Reactor Protection and ESF Logic Actuation - Logic Test, Revision 64
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
On February 22, 2011, the inspectors reviewed and observed the performance of an emergency planning drill that involved a turbine runback, loose parts alarms for the reactor vessel, loss of an emergency electrical bus, a steam generator tube rupture, an overspeed trip of the turbine driven auxiliary feedwater pump, and a failed open main steam relief valve resulting in an Alert and subsequent Site Area Emergency, followed by a General Emergency. The inspectors assessed emergency procedure usage, emergency plan classification, notifications, and the licensees identification and entrance of any problems into their corrective action program. This inspection evaluated the adequacy of the licensees conduct of the drill and critique performance. The licensee captured drill deficiencies within their corrective action program (CAP).
b. Findings
No findings were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
a. Inspection Scope
The inspectors performed a periodic review of the three following Unit 1 and 2 PIs to assess the accuracy and completeness of the submitted data and whether the performance indicators were calculated in accordance with the guidance contained in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6. The inspection was conducted in accordance with NRC inspection procedure 71151, Performance Indicator Verification. Specifically, the inspectors reviewed the Unit 1 and Unit 2 data reported to the NRC for the period January 1, 2010, through December 31, 2010. Documents reviewed included applicable NRC inspection reports, licensee event reports, operator logs, station performance indicators, and related CRs, specifically CR 401933, Re-evaluate PI for 5/28/2010 reactor trip with loss of 2H emergency bus.
- Unplanned Scrams per 7000 Critical Hours
- Unplanned Transients per 7000 Critical Hours
- Unplanned Scrams With Complications
b. Findings
No findings were identified.
4OA2 Identification and Resolution of Problems
Review of Items Entered into the Corrective Action Program
a. Inspection Scope
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 CAP. This review was accomplished by reviewing daily CR report summaries and periodically attending daily CR Review Team meetings.
b. Findings
No findings were identified.
4OA3 Event Followup
.1 (Closed) Licensee Event Report (LER) 05000339/2010-004-00:
Automatic Reactor Trip and Engineered Safety Feature Actuation Due to Lightning Strike
a. Inspection Scope
On June 16, 2010, with Unit 2 operating in Mode 1 at 98% power, an automatic reactor trip occurred caused by an Overtemperature Delta T (OTdT) reactor trip signal without an actual overtemperature condition being present. The direct cause of the OTdT reactor trip signal was a lightning strike which induced a voltage transient of sufficient magnitude on two channels of the OTdT protection circuits to cause actuation of the reactor protection system which resulted in a Unit 2 reactor trip. This LER is closed.
b. Findings
Introduction:
A non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," was identified by the inspectors for the licensees failure to determine the cause of a significant condition adverse to quality involving an automatic reactor trip following a lightning strike on the Unit 2 containment building. This resulted in the Unit 2 automatic reactor trip on June 16, 2010, because of the insufficient corrective action to preclude repetition.
Description:
On June 16, 2010, Unit 2 experienced an automatic reactor trip during a thunderstorm in which lightning struck the Unit 2 containment building. The automatic reactor trip signal originated from Channels 1 and 2 of the OTdT protective circuitry. The licensee entered this issue into their CAP as CR384967, U-2 tripped the evening of June 16, 2010, during a severe thunderstorm and initiated RCE001015, Reactor trip due to lightning strike. The inspectors reviewed RCE001015 and CR384967. The inspectors noted that the root cause of the event was determined to be the absence of a lag function in the OTdT protective circuitry that had been removed in 1995 as part of a reactor coolant system (RCS) resistance temperature detector (RTD) bypass line removal design change. RCE001015 also documented the improper grounding of an RCS cold leg wide range temperature RTD cable shield as a contributing cause. This improper grounding occurred during one of a series of Appendix R design changes that were completed between 1983 and 1985.
The inspectors noted that, in August 2005 Unit 2 also experienced an automatic reactor trip during a thunderstorm in which lightning struck the Fuel Building. The automatic reactor trip signal originated from Channels 1 and 2 of the OTdT protective circuitry. The inspectors reviewed the licensees investigation, RCE N-2005-3016, Unit 2 Reactor Trip on Overtemperature Delta Temperature, which determined the root cause to be improperly grounded shields for the Unit 2 Channels 1 and 2, RCS hot and cold leg, spare RTDs. Corrective action to prevent recurrence taken as a result of this root cause was to correct the improper spare RTD cable grounding.
The inspectors noted that at the time of the 2005 event the licensee implemented their CAP via procedures such as, VPAP-1601, Corrective Action, Revision 20 and DNAP-1604, Cause Evaluation Program, Revision 1. The inspectors reviewed the licensees CAP relative to the treatment of the August 2005 reactor trip and noted the following:
- VPAP-1601, section 4.23 defines Significant, in part, as: Events that need immediate attention to prevent recurrence.
- VPAP-1601, Attachment 1 identifies reactor trips as Significant.
- DNAP-1604, section 3.2.10 directs that for RCEs, to Determine Corrective Actions to Prevent Recurrence (CAPRs) for each root cause.
The inspectors concluded that the reactor trip from critical conditions that occurred in 2005 during on site lightning strikes was a SCAQ. The inspectors also concluded that the licensee failed to determine the cause of the SCAQ, and as a result failed to take action to preclude repetition.
Analysis:
The inspectors determined that the failure to determine the cause of a SCAQ was a performance deficiency. The inspectors reviewed IMC 0612, Appendix B and determined the performance deficiency was more than minor because, if left uncorrected, it has the potential to lead to a more significant safety concern in that failing to identify the cause of SCAQs and thus failing to take corrective action to preclude repetition could result in additional initiating events or impacts on mitigating systems. In addition, the inspectors determined that it adversely impacted the Initiating Events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, specifically the attribute of Protection Against External Factors in that the removal of the OTdT lag function removed protection from lightning strikes on the reactor protection system. The inspectors reviewed IMC 0609, Attachment 4 and determined that the finding was of very low safety significance, or Green, because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The cause of this finding involved the cross-cutting area of problem identification and resolution, the component of operating experience, and the aspect of evaluation of identified problems, P.1(c), because the licensee failed to thoroughly evaluate the cause of the 2005 reactor trip and conduct effectiveness reviews of corrective actions to ensure the problems are resolved.
Enforcement:
10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that measures shall be established to assure that the cause of significant conditions adverse to quality are identified and corrective action taken to preclude repetition.
Contrary to the above, on June 16, 2010, the licensee failed to determine the cause of a significant condition adverse to quality, and as a result failed to take action to preclude repetition. Because the finding is of very low safety significance (Green) and it was entered into the licensees CAP as CR384967, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000339/2011002-03, Failure to Determine the Cause and Take Corrective Action to Preclude Repetition for Lightning Induced Reactor Trips.
.2 (Closed) Licensee Event Report (LER) 05000338/2010-004-00:
Manual Reactor Trip due to Malfunction of the Rod Control In-Hold-Out Selector Switch
On October 22, 2010, Unit 1 was in Mode 2 with zero power physics testing (ZPPT) in progress in accordance with licensee procedure 1-PT-94.0, Refueling Nuclear Design Check Tests, Revision 32, when the licensee made the decision to perform a manual reactor trip in order to repair the rod control in-hold-out selector switch which is non-safety related. The switch was subsequently found with dirty contacts leading to sluggish operation and replaced. The licensee entered this problem in their CAP as CR400145. This LER is closed.
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 the licensee security procedures and regulatory requirements relating to nuclear plant security.
These observations took place during both normal and off-normal plant working hours.
These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors normal plant status review and inspection activities.
b. Findings
No findings were identified.
4OA6 Meetings, Including Exit
.1
Exit Meeting Summary
On April 21, 2011, the senior resident inspector presented the inspection results to Mr.
Fred Mladen and other members of the staff, who acknowledged the findings. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.
On April 25, 2011, the senior resident inspector performed a phone re-exit with Mr. Don Taylor who acknowledged the findings. On May 2, 2011, the senior resident inspector performed a phone re-exit with Mr. Don Taylor who acknowledged the findings.
ATTACHMENT: SUPPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- W. Anthes, Manager, Nuclear Maintenance
- M. Becker, Manager, Nuclear Outage and Planning
- M. Crist, Plant Manager
- R. Evans, Manager, Radiological Protection and Chemistry
- T. Huber, Director, Nuclear Engineering
- S. Hughes, Manager, Nuclear Operations
- C. Gum, Manager, Nuclear Protection Services
- L. Lane, Site Vice President
- J. Lieberstien, Technical Advisor, Licensing
- P. Kemp, Manager, Organizational Effectiveness
- F. Mladen, Director, Station Safety and Licensing
- B. Morrison, Supervisor Nuclear Engineering
- C. McClain, Manager, Nuclear Training
- R. Scanlon, Manager, Nuclear Site Services
- J. Scott, Supervisor, Nuclear Training (operations)
- D. Taylor, Supervisor, Station Licensing
- M. Whalen, Technical Advisor, Licensing
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
None
Opened and Closed
Inadequate Installation of Unit 2 Low Head Safety
Injection Piping and Related Supports (Section 1R15)
05000338, 339/2011002-02 NCV
Inadequate Post Maintenance Test Program Instructions
for Safety-Related Instrument and Control Preventative
Maintenance (Section 1R19)05000339/2011002-03
Failure to Determine the Cause and Take Corrective
Action to Preclude Repetition for Lightning Induced
Reactor Trips (Section 4OA3.1)
Closed
05000339/2010-004-00
LER
Automatic Reactor Trip and Engineered Safety Feature
Actuation Due to Lightning Strike (Section 4OA3.1)
05000338/2010-004-00 LER Manual Reactor Trip due to Malfunction of the Rod Control
In-Hold-Out Selector Switch (Section 4OA3.2)
Discussed
None
LIST OF ACRONYMS
Agencywide Document Access and Management System
CA
Corrective Action
Corrective Action Program
CFR
Code of Federal Regulations
CR
Condition Report
IMC
Inspection Manual Chapter
LHSI
Low Head Safety Injection
Non-cited Violation
NRC
Nuclear Regulatory Commission
Publicly Available Records
Performance Indicator
QS
Quench Spray
Root Cause Evaluation
Reactor Coolant Pump
Rated Thermal Power
Significance Determination Process
SR
Surveillance Requirements
TDAFWP
Turbine Driven Auxiliary Feedwater Pump
TS
Technical Specifications
Updated Final Safety Analysis Report
Unresolved Item
Virginia Electric and Power Company
VPAP
Virginia Power Administrative Procedure
Work Order