IR 05000271/2013002
| ML13109A505 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 04/19/2013 |
| From: | Bellamy R NRC/RGN-I/DRP/PB5 |
| To: | Wamser C Entergy Nuclear Operations |
| Bellamy R | |
| References | |
| IR-13-002 | |
| Download: ML13109A505 (43) | |
Text
UNITED STATES April 19, 2013
SUBJECT:
VERMONT YANKEE NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000271/2013002
Dear Mr. Wamser:
On March 31, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Vermont Yankee Nuclear Power Station. The enclosed inspection report documents the inspection results, which were discussed on April 15, 2013, with you 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.
This report documents three NRC identified findings of very low safety significance (Green).
These findings were determined to involve violations of NRC requirements. Additionally, this report documents the final resolution for the self-revealing apparent violation associated with the failure of the B emergency diesel generator documented in the 4th quarter 2012 integrated inspection report, 05000271/2012005. This apparent violation was also determined to be a finding of very low safety significance (Green). However, because of the very low safety significance, and because they are entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs), consistent with Section 2.3.2 of the NRC Enforcement Policy. If you contest any NCVs in this report, 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 I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Vermont Yankee.
If you disagree with the cross-cutting aspect assigned to 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 I, and the NRC Resident Inspector at Vermont Yankee.
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 Agencywide Document Access and Management 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/
Ronald R. Bellamy, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket No. 50-271 License No. DPR-28
Enclosure:
Inspection Report No. 05000271/2013002 w/ Attachment: Supplementary Information
REGION I==
Docket No.: 50-271 License No.: DPR-28 Report No.: 05000271/2013002 Licensee: Entergy Nuclear Operations, Inc.
Facility: Vermont Yankee Nuclear Power Station Location: Vernon, Vermont 05354-9766 Dates: January 1, 2013 through March 31, 2013 Inspectors: S. Rutenkroger, PhD, Senior Resident Inspector, Division of Reactor Projects (DRP)
S. Rich, Resident Inspector, DRP J. DeBoer, Acting Resident Inspector, DRP R. Nimitz, Senior Health Physicist, Division of Reactor Safety (DRS)
J. Furia, Senior Health Physicist, DRS J. Laughlin, Emergency Preparedness Inspector, Office of Nuclear Security and Incident Response Approved by: Ronald R. Bellamy, PhD, Chief Reactor Projects Branch 5 Division of Reactor Projects Enclosure
SUMMARY
IR 05000271/2013002; 01/01/2013 - 03/31/2013; Vermont Yankee Nuclear Power Station;
Equipment Alignment, Plant Modifications, Problem Identification and Resolution and Other Activities.
This report covered a three-month period of inspection by resident inspectors and announced inspections performed by regional inspectors. There were three NRC-identified findings and one self-revealing finding of very low safety significance (Green), all of which were non-cited violations (NCVs), documented in this report. The significance of inspection findings are indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP),
dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Components Within the Cross-Cutting Areas, dated October 28, 2011. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated January 28, 2013.
The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified an NCV of operating license condition 3.F, fire protection program, because Entergy did not correct a degraded latch on a three-hour rated fire door on the entrance to the B emergency diesel generator (EDG) room, and as a result the three-hour fire barrier was non-functional and the required compensatory measure of an hourly fire watch was not in effect. Entergys corrective actions included restoring vertical alignment of the latching mechanism, further inspection by a locksmith to ensure reliable operation, planning a preventive replacement of the latch due to existing excessive wear, and initiating a condition report.
The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the fire door being degraded with unreliable latching without an assigned hourly fire watch from January 20 to January 22 resulted in a barrier to fire propagation that was less robust than required by the approved fire protection program. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that this finding is of very low safety significance (Green) per Task 1.3.2, Task 1.3.2: Supplemental Screening for Fire Confinement Findings. The inspectors determined the degradation rating associated with the deficiency to be Moderate B since a closure mechanism held the door against the door jamb, the door swings out from the EDG room, no combustibles were stored in the adjacent hallway, and no equipment important to safety exists in the turbine building hallway. Therefore, the degraded fire door provided a minimum of 20 minutes of fire endurance protection and the fixed or in situ fire ignition sources and combustible or flammable materials were positioned such that, even considering fire spread to secondary combustibles, the degraded fire door would not have been subject to direct flame impingement since no combustible material was located near the door during the time of concern. The inspectors determined that the finding had a cross-cutting aspect in the Problem Identification and Resolution area, Corrective Action Program component, because Entergy personnel did not completely identify the issue with the alignment of the striker plate when the degradation was first identified and did not identify that the latching deficiency still existed during subsequent transits through the door P.1(a).
(Section 1R04)
- Green.
The inspectors identified an NCV of Technical Specification 6.4, Procedures, because procedure OPOP-SW-2181, Service Water/ Alternate Cooling System, was inadequate. Specifically, the step in the procedure to identify and isolate sources of water lost from the cooling tower basin would not have been implemented in a timely manner while a temporary fire water system was drawing on the basin. Entergys corrective actions included writing a night order describing the fire fighting strategy for a fire in the intake and directing the temporary fire pumps to be stopped if they started automatically while the alternate cooling system (ACS) was in service, implementing temporary procedure changes, and initiating a condition report.
The finding is more than minor because it impacted the design control attribute of the Mitigating Systems cornerstone. Specifically, the temporary modification added another potential path for loss of water from the cooling tower deep basin and the appropriate compensatory measures to address that loss path were not implemented, impacting the capability and reliability of ACS. Additionally, the finding is similar to IMC 0612, Appendix E,
Examples of Minor Issues, example 3.j more than minor description, because the added draw on the cooling tower basin water had the potential to affect the accident analysis calculation assumption of the amount of water available for running ACS. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not involve the total loss of a safety function that contributes to external event initiated core damage accident sequences. This condition existed for less than the technical specification allowed outage time of seven days. This finding had a cross-cutting aspect in the area of human performance, Work Control, because Entergy did not appropriately coordinate work activities by incorporating actions to address the need to keep personnel apprised of the operational impact of work activities. Specifically, Entergy identified the need for compensatory measures for the temporary modification for the fire water system work, but the necessary actions were not coordinated to ensure operations and maintenance understood the operational impact of the work H.3(b). (Section 1R18)
- Green.
The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI,
Corrective Action, because Entergy did not promptly correct an adverse condition resulting in the failure of the B-UPS-1A low pressure coolant injection (LPCI) uninterruptible power supply (UPS) battery. Specifically, Entergy personnel did not promptly replace a degraded battery cell prior to its exceeding operability limits. Entergys corrective actions included replacing cell 61, replacing all cells with individual cell voltages (ICVs) less than 2.13 V, expediting complete battery bank replacements with a due date of May 30, and initiating a condition report.
The finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, UPS-1A had unplanned inoperability and degraded capacity due to cell 61 being out of service which commenced at some unknown point between December 3 and December 9 and was restored when cell 61 was replaced on December 10. In accordance with IMC 0609.04, Initial Characterization of
Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because it did not represent a loss of system safety function or a loss of safety function for a single train (UPS-1A and A LPCI) for greater than its technical specification allowed outage time (seven days). The inspectors determined that the finding had a cross-cutting aspect in the Problem Identification and Resolution area, Operating Experience component, because Entergy personnel did not implement and institutionalize available operating experience guidance contained within IEEE-450, IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications, or alternatively, vendor recommendations, to support plant safety P.2(b).
(Section 4OA2)
- Green.
A self-revealing NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified because Entergy did not promptly correct an adverse condition resulting in the failure of the B EDG. Specifically, Entergy personnel did not promptly replace a degraded jacket water flange gasket prior to its subsequent failure. Entergys corrective actions included replacing the gasket, visually inspecting the other jacket water connections, and initiating a condition report.
The finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the B EDG failed in service due to a known degraded condition that affected the overall system redundancy and reliability and resulted in 37 days of unplanned unavailability. The inspectors and a Region I Senior Reactor Analyst (SRA) completed the Detailed Risk Evaluation (DRE) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined this finding to be of very low safety significance (Green). The DRE estimated the increase in core damage frequency (CDF) for internal initiating events in the range of 1 core damage accident in 2,000,000 years of reactor operation, in the mid-E-7 range per year. In addition, external initiating events such as fire, seismic and flooding would not have increased the total CDF above 1 E-6 per year, and the increase in the frequency of a large early release of radioactive material (LERF) associated with the internal event CDF core damage sequences would not be above 1E-7 per year. The finding had a cross-cutting aspect in the Human Performance, Decision-Making, because Entergy personnel did not use conservative assumptions in decision making in that the chosen action was to monitor the leak for a prolonged period of time H.1(b). (Section 4OA5)
REPORT DETAILS
Summary of Plant Status
Vermont Yankee Nuclear Power Station (VY) began the inspection period operating at 100 percent power. On January 14, operators reduced power to 52 percent for a control rod pattern adjustment and returned VY to 100 percent power the following day. On January 17, operators reduced power to 76 percent for a control rod pattern adjustment and returned VY to 100 percent power the same day. On March 9, operators shut down the reactor to conduct a refueling outage. The plant remained shut down for the refueling outage for the remainder of the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01 Adverse Weather Protection
Readiness for Impending Adverse Weather Conditions
a. Inspection Scope
The inspectors reviewed Entergys preparations for the onset of winter storm weather, including high winds, on February 8. The inspectors reviewed the implementation of adverse weather preparation procedures before the onset of this adverse weather condition. The inspectors walked down outside areas, the EDGs, intake structure, and service water pumps to ensure system availability. The inspectors verified that operator actions defined in Entergys adverse weather procedure maintained the readiness of essential systems. The inspectors discussed readiness and staff availability for adverse weather response with operations and work control personnel.
b. Findings
No findings were identified.
1R04 Equipment Alignment
.1 Partial System Walkdowns
a. Inspection Scope
The inspectors performed partial walkdowns of the following systems:
A EDG during B service water pump maintenance on January 22 Service water pumps and service water strainers during cooling tower cell 2-1 maintenance on February 4 B shutdown cooling while A shutdown cooling was unavailable on March 13 The inspectors selected these systems based on their risk-significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors reviewed applicable operating procedures, system diagrams, the Updated Final Safety Analysis Report (UFSAR), technical specifications (TS), condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted system performance of their intended safety functions. The inspectors also performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and were operable.
The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. The inspectors also reviewed whether Entergy staff had properly identified equipment issues and entered them into the corrective action program for resolution with the appropriate significance characterization.
b. Findings
Introduction.
The inspectors identified a Green NCV of operating license condition 3.F, fire protection program, because Entergy did not implement and maintain in effect all provisions of the NRC approved fire protection program. Specifically, Entergy did not correct a degraded latch on a three-hour rated fire door on the entrance to the B EDG room and as a result the three hour fire barrier was non-functional and the required compensatory measure of an hourly fire watch was not in effect.
Description.
On January 20, Entergy personnel identified that the door on the entrance to the B EDG room, Appendix R classified fire door ACD-116-303, was not latched closed. Entergy maintenance personnel determined that the latch mechanism was not latching reliably due to being degraded. Entergy declared the door non-functional and implemented an hourly fire watch in accordance with the Technical Requirements Manual (TRM) 3.13.E.2, and fire protection plan. Maintenance personnel made minor adjustments and tested the door by closing and latching it successfully multiple times, and Entergy declared the door functional and stopped the hourly fire watch.
On January 22, the inspectors identified that the door, although appearing to be latched closed, did not latch when it was shut. Entergy declared the door non-functional and implemented an hourly fire watch in accordance with the TRM and fire protection plan.
Maintenance personnel identified that the latch mechanism required significant adjustment in order to address an alignment issue and shimmed the striker plate to restore vertical alignment.
The inspectors determined that the maintenance performed on January 20 was insufficient to ensure continued reliable door operation. In addition, the inspectors determined that Entergy personnel had reasonable opportunities to identify the further latch degradation due to frequent rounds performed by Entergy personnel that transit through the door each shift.
Entergys corrective actions included restoring vertical alignment of the latching mechanism, further inspection by a locksmith to ensure reliable operation, planning a preventive replacement of the latch due to existing excessive wear, and initiating CR-VTY-2013-00394.
Analysis.
The inspectors determined that Entergy personnels failure to initially correct the non-functional fire door and failure to identify the continued latching deficiency was a performance deficiency that was reasonably within Entergys ability to foresee and correct and should have been prevented. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the fire door being degraded with unreliable latching without an assigned hourly fire watch from January 20 to January 22 resulted in a barrier to fire propagation that was less robust than required by the approved fire protection program.
In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that this finding is of very low safety significance (Green) per Task 1.3.2, Task 1.3.2: Supplemental Screening for Fire Confinement Findings. The inspectors determined the degradation rating associated with the deficiency to be Moderate B since a closure mechanism held the door against the door jamb, the door swings out from the EDG room, no combustibles were stored in the adjacent hallway, and no equipment important to safety exists in the turbine building hallway. Therefore, the degraded fire door provided a minimum of 20 minutes of fire endurance protection and the fixed or in situ fire ignition sources and combustible or flammable materials were positioned such that, even considering fire spread to secondary combustibles, the degraded fire door would not have been subject to direct flame impingement since no combustible material was located near the door during the time of concern.
The inspectors determined that the finding had a cross-cutting aspect in the Problem Identification and Resolution area, Corrective Action Program component, because Entergy personnel did not completely identify the issue with the alignment of the striker plate when the degradation was first identified and did not identify that the latching deficiency still existed during subsequent transits through the door P.1(a).
Enforcement.
License condition 3.F requires, in part, that Entergy shall implement and maintain in effect all provisions of the approved fire protection program. A provision of the approved fire protection program is maintaining a qualified three hour fire door, ACD-116-303, in accordance with TRM 3.13.E.2. Contrary to the above, from January 20 to January 22, 2013, fire door ACD-116-303 had an unreliable latching mechanism, rendering it a non-functional three hour fire barrier, and the actions required by TRM 3.13.E.2 to conduct an hourly fire watch were not performed. Because this violation was of very low safety significance (Green), and Entergy entered this issue into their corrective action program (CR-VTY-2013-00394), this violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. (NCV
===05000271/2013002-01, Appendix R Fire Door Not Latching Closed Due to Misalignment)
.2 Full System Walkdown
a. Inspection Scope
Between March 1 and 5, the inspectors performed a complete system walkdown of accessible safety-related portions of the high pressure coolant injection system to verify the existing equipment lineup was correct. The inspectors reviewed operating procedures, drawings, equipment line-up check-off lists, recent condition reports, the system health report and the UFSAR to verify the system was aligned to perform its required safety functions. The inspectors also reviewed electrical power availability, component lubrication, hangar and support functionality, and operability of support systems. The inspectors performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies.
Additionally, the inspectors reviewed a sample of related condition reports to ensure Entergy appropriately evaluated and resolved any deficiencies. The inspectors discussed the systems condition with the system engineer.
b. Findings
No findings were identified.
1R05 Fire Protection
Resident Inspector Quarterly Walkdowns (71111.05Q - 5 samples)
a. Inspection Scope
The inspectors conducted tours of the areas listed below to assess the material condition and operational status of fire protection features. The inspectors verified that Entergy controlled combustible materials and ignition sources in accordance with administrative procedures. The inspectors verified that fire protection and suppression equipment was available for use as specified in the area pre-fire plan, and passive fire barriers were maintained in good material condition. The inspectors also verified that station personnel implemented compensatory measures for out of service, degraded, or inoperable fire protection equipment, as applicable, in accordance with procedures.
East and West switchgear rooms on January 28 Reactor building 303 elevation on January 30 Reactor building 318 elevation on January 30 Turbine building condenser bay and ground floor 282 6 elevation on March 17 Primary containment on March 18
b. Findings
No findings were identified.
1R06 Flood Protection Measures
Internal Flooding Review
a. Inspection Scope
The inspectors reviewed the UFSAR, the site flooding analysis, and drawings to assess susceptibilities involving internal flooding. The inspectors also reviewed the corrective action program to determine if Entergy identified and corrected flooding problems and whether operator actions for coping with flooding were adequate. The inspectors focused on the Reactor Building 303 and 318 elevations to verify the adequacy of equipment seals located below the flood line, floor and water penetration seals, common drain lines and sumps.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program
.1 Quarterly Review of Licensed Operators Requalification Testing and Training
a. Inspection Scope
The inspectors observed licensed operator simulator training on March 28 which included achieving criticality, establishing a heat up rate, safety/relief valve testing and recovering from an excessive heat up rate. The inspectors evaluated operator performance during the start up just in time training and verified completion of risk significant operator actions. The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and changing plant conditions, and the oversight and direction provided by the control room supervisor.
Additionally, the inspectors assessed the ability of the crew and training staff to identify and document crew performance problems.
b. Findings
No findings were identified.
.2 Quarterly Review of Licensed Operator Performance in the Main Control Room
a. Inspection Scope
The inspectors observed control room operators during a planned down power on January 14 for a control rod pattern sequence exchange. The inspectors observed the pre-job briefings to verify that roles and responsibilities, critical steps, expected results and hold points were discussed. The inspectors verified that procedure use, crew communications, and response to alarms met established expectations and standards.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed the samples listed below to assess the effectiveness of maintenance activities on structure, system, and component (SSC) performance and reliability. The inspectors reviewed system health reports, corrective action program documents, maintenance work orders, and maintenance rule basis documents to ensure that Entergy was identifying and properly evaluating performance problems within the scope of the maintenance rule. For each sample selected, the inspectors verified that the SSC was properly scoped into the maintenance rule in accordance with 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, and verified that the (a)(2) performance criteria established by Entergy staff were reasonable. Additionally, the inspectors ensured that Entergy staff was identifying and addressing common cause failures that occurred within and across maintenance rule system boundaries.
Fire water pumps Reactor protection system
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed station evaluation and management of plant risk for the maintenance and emergent work activities listed below to verify that Entergy performed the appropriate risk assessments prior to removing equipment for work. The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that Entergy personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When Entergy performed emergent work, the inspectors verified that operations personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work and discussed the results of the assessment with the stations work week manager to verify plant conditions were consistent with the risk assessment. The inspectors also reviewed the TS requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.
A EDG monthly surveillance, A LPCI unavailable during torus cooling, and A residual heat removal (RHR) and residual heat removal service water (RHRSW)quarterly surveillances - workweek (WW) 1301 B service water pump maintenance, C reactor feedwater pump minimum flow line isolated, B RHR and RHRSW quarterly surveillances, and UPS-1A maintenance -
WW 1304 A EDG monthly surveillance, B service water pump corrective maintenance, and C service water pump maintenance - WW 1305 Cooling tower cell 2-1 maintenance, A service water pump maintenance, and reactor core isolation cooling (RCIC) actuation logic surveillance - WW 1306 A EDG maintenance, unit auxiliary transformer replacement, and electrical bus maintenance during refueling outage (RFO) 30 Reactor building blowout panel unplanned displacement and modified replacement panel installation during RFO 30
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments
a. Inspection Scope
The inspectors reviewed operability determinations for the following degraded or non-conforming conditions:
Trip setpoint for a level transmitter for low reactor water level was found outside the range required by the calibration procedure, condition report CR-VTY-2013-00104 initiated on January 7 B service water pump breaker was found to be missing an internal support bolt, condition report CR-VTY-2013-00448 initiated on January 24 Silicone foam seal between the turbine lube oil room and heater bay was found to have a crack in the exposed surface, condition report CR-VTY-2013-00555 initiated on January 30 RCIC inboard steam line drain isolation valve took seven minutes to indicate full closed during the quarterly surveillance, condition report CR-VTY-2013-00878 initiated on February 14 B EDG was found to have a jacket water cooling leak of 70 drops per minute from the floating channel head of the heat exchanger, condition report CR-VTY-2013-01317 initiated on March 9 The inspectors selected these issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the operability determinations to assess whether TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS and UFSAR to Entergys evaluations to determine whether the components or systems were operable. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations.
b. Findings
No findings were identified.
1R18 Plant Modifications
.1 Temporary Modifications
a. Inspection Scope
The inspectors reviewed a temporary modification to install replacement fire water pumps that drew water from the cooling tower basin to determine whether the modification affected the safety functions of systems that are important to safety. The inspectors reviewed the process applicability determination documentation and post-modification testing results, and conducted field walkdowns of the modification to verify that the temporary modification did not degrade the design bases, licensing bases, and performance capability of the affected systems.
b. Findings
Introduction.
The inspectors identified a Green NCV of TS 6.4, Procedures, because procedure OPOP-SW-2181, Service Water/ Alternate Cooling System, was inadequate. Specifically, the step in the procedure to identify and isolate sources of water loss from the cooling tower basin would not have been implemented in a timely manner while a temporary fire water system was drawing on the basin.
Description.
On February 20, Entergy completed installing a temporary modification that would provide water for the fire suppression system in order to allow the normal fire water pumps to be taken out of service for system maintenance. The temporary modification included two diesel-driven pumps that provided a flow of 2500 gallons per minute each and a smaller, constantly running, jockey pump. The diesel driven pumps would start automatically once fire header pressure dropped below a pressure set point, regardless of the cause of the low pressure. For example, the pumps would start given actuation of a sprinkler system, use of a fire hose, or a break in the non-safety related piping that makes up the fire water system. A pump operator stationed by the pump would notify the control room that a pump had started and would also start or stop the pumps as directed.
The temporary fire water supply system drew water from the deep basin underneath cooling tower 2. This deep basin is the safety related ultimate heat sink in the event of a loss of the intake structure, and can be placed in service by performing the valve line-up in procedure OPOP-SW-2181 in less than two hours. Flow from the service water pumps, located in the intake structure, constantly maintains the basin full so that the ACS is operable and available to respond to a loss of intake (i.e. the service water pumps).
When designing the temporary modification, Entergy personnel recognized that ACS could be impacted by an event that caused a loss of all service water and caused the temporary fire water pumps to start, and specified compensatory measures to reduce the likelihood of a large fire in the intake, as well as the unnecessary operation of the temporary pumps if ACS was needed. A continuous fire watch in the intake structure was required because a fire in that location could cause a loss of the service water pumps. The FM200 fire suppression system for the diesel fire pump fuel oil day tank was required to be in service because the fuel oil tank is a large source of combustible material in the intake structure. The engineering change also stated that operators would be directed to secure the temporary fire pumps if they started while ACS was in use. These were necessary because OPOP-SW-2181 did not direct identifying and isolating nonessential uses of the basin water until after the lineup was mostly complete, between one and two hours into the event that caused a loss of all service water.
On February 21, the inspectors questioned the control room staff on the impact of the temporary modification on ACS, and their actions if the temporary fire water pump(s)were to start while ACS was in use. The operators reviewed the engineering change and identified that the compensatory actions described above were not implemented.
Specifically, there was only an hourly fire watch in the intake structure, the FM 200 system was tagged out, and there was no direction on the use of the temporary fire pumps while ACS was in service. In addition, the inspectors asked to review the temporary procedure changes made to reflect the unavailability of the normal fire water pumps and found that the changes had not been implemented.
Entergy took immediate actions to change the hourly fire watch to continuous, to return the FM200 system to service, and to write a night order describing the fire fighting strategy for a fire in the intake and directing the temporary fire pumps to be stopped if they started automatically while ACS was in service. They also implemented the temporary procedure changes. The time between when the temporary modification was able to drain the basin and when the compensatory measures were implemented was less than the seven day TS allowed outage time for ACS. Entergy initiated condition report CR-VTY-2013-01001.
Entergy determined that with the basin full, there was an available margin of 147,991 gallons beyond what was needed for ACS to operate for seven days per its design. If flow to the cooling tower deep basin from the service water pumps was lost, and one of the temporary fire water pumps was running, it would take about an hour for the basin level to drop below the level required to maintain ACS operable. Entergy determined that this was a reasonable amount of time for the operators to identify the impact on the deep basin and direct the pump operator to stop the fire pump even without proceduralized guidance. Based on this, Entergy determined that ACS was maintained available. The inspectors determined that the addition of equipment removing water from the deep basin automatically that required manual operator action to preserve functionality of ACS without specific written procedures in place to preserve the safety related function of ACS rendered ACS inoperable.
Analysis.
The inspectors determined that Entergys failure to implement the compensatory measures specified in the engineering change package was within their ability to foresee and correct and should have been prevented. This performance deficiency is more than minor because it impacted the design control attribute of the Mitigating Systems cornerstone. Specifically, the temporary modification added another potential path for loss of water from the cooling tower deep basin and the appropriate compensatory measures to address that aspect were not implemented, impacting the capability and reliability of ACS. Additionally, the finding is similar to IMC 0612, Appendix E, Examples of Minor Issues, example 3.j. more than minor description, because the added draw on the cooling tower basin water had the potential to affect the accident analysis calculation assumption of the amount of water available for running ACS.
In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not involve the total loss of a safety function that contributes to external event initiated core damage accident sequences. This condition existed for less than the TS allowed outage time of seven days.
This finding had a cross-cutting aspect in the area of human performance, Work Control, because Entergy did not appropriately coordinate work activities by incorporating actions to address the need to keep personnel apprised of the operational impact of work activities. Specifically, Entergy identified the need for compensatory measures for the temporary modification for the fire water system work, but the necessary actions were not coordinated to ensure operations and maintenance personnel understood the operational impact of the work. H.3(b)
Enforcement.
TS 6.4, Procedures, requires that written procedures be established, implemented and maintained covering operation of systems and components. Contrary to this, procedure OPOP-SW-2181, Service Water/ Alternate Cooling System was not adequately maintained. Specifically, the step in the procedure to identify and isolate sources of water loss from the cooling tower basin would not have been implemented in a timely manner while a temporary fire water system was drawing on the basin.
Entergys immediate corrective actions included implementing the required compensatory measures and initiating condition report CR-VTY-2013-01001. Because this violation was of very low safety significance (Green), and Entergy entered this issue into their corrective action program, this violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. (NCV 05000271/2013002-02, Failure to Implement Compensatory Measures Associated with a Temporary Modification)
.2 Permanent Modifications
a. Inspection Scope
The inspectors evaluated an engineering change for the reactor building air recirculation cooling units (RRU) that involved changing from corrected differential pressure tests to a thermal performance test to support operability of the units. The inspectors verified that the design bases, licensing bases, and performance capability of the RRUs were not degraded by the modification of the new testing procedures for operability. The inspectors reviewed modification documents associated with the design change and the inspectors also interviewed engineering and maintenance personnel involved with the modification.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed the post-maintenance tests for the maintenance activities listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the test procedure to verify that the procedure adequately tested the safety functions that may have been affected by the maintenance activity, that the acceptance criteria in the procedure were consistent with the information in the applicable licensing basis and/or design basis documents, and that the procedure had been properly reviewed and approved. The inspectors also witnessed the test or reviewed test data to verify that the test results adequately demonstrated restoration of the affected safety functions.
D service water pump cable replacement, test on January 17 B service water pump cable replacement, test on January 24 B EDG cylinder liner replacements and maintenance, testing on March 25 to 28 Following repair and replacement fire water valves - diesel fire pump discharge valve, electric fire pump discharge valve, and diesel fire pump discharge to test stand valve, test on March 1 B normal fuel pool cooling pump maintenance, test on February 28 Emergent secondary containment repairs due to ventilation system overpressure, test on March 20
b. Inspection Scope
No findings were identified.
1R20 Refueling and Other Outage Activities
a. Inspection Scope
The inspectors reviewed the stations work schedule and outage risk plan for RFO 30, which began March 9. The inspectors reviewed Entergys development and implementation of outage plans and schedules to verify that risk, industry experience, previous site-specific problems, and defense-in-depth were considered. During the outage, the inspectors observed portions of the shutdown and cooldown processes and monitored controls associated with the following outage activities:
Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable TSs when taking equipment out of service Implementation of clearance activities and confirmation that tags were properly hung and that equipment was appropriately configured to safely support the associated work or testing Installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication and instrument error accounting Status and configuration of electrical systems and switchyard activities Monitoring of decay heat removal operations Impact of outage work on the ability of the operators to operate the spent fuel pool cooling system Reactor water inventory controls, including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss Activities that could affect reactivity Maintenance of secondary containment as required by TSs Refueling activities, including fuel handling and fuel receipt inspections Fatigue management Identification and resolution of problems related to refueling outage activities
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed performance of surveillance tests and reviewed test data of selected risk-significant SSCs to assess whether test results satisfied TS, the UFSAR, and Entergys procedure requirements. The inspectors verified that test acceptance criteria were clear, tests demonstrated operational readiness and were consistent with design documentation, test instrumentation had current calibrations and the range and accuracy for the application, tests were performed as written, and applicable test prerequisites were satisfied. Upon test completion, the inspectors considered whether the test results supported that equipment was capable of performing the required safety functions. The inspectors reviewed the following surveillance tests:
Reactor building air recirculation cooling unit RRU-7, thermal performance test on January 3 A and C RHRSW quarterly surveillance on January 3 (in-service test)
Seismic monitoring system quarterly surveillance on February 5 Delayed access power source backfeed once per operating cycle surveillance on March 10 B main steam isolation valve local leak rate test (LLRT) on March 13 (containment isolation valve)
RCIC steam supply line isolation valve LLRT on March 13 (containment isolation valve)
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes
a. Inspection Scope
The Office of Nuclear Security and Incident Response headquarters staff performed an in-office review of the latest revisions of various Emergency Plan Implementing Procedures and the Emergency Plan located under ADAMS accession numbers ML13003A141, ML13003A142 and ML130230023.
Entergy determined that in accordance with 10 CFR 50.54(q), the changes made in the revisions resulted in no reduction in the effectiveness of the Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, this revision is subject to future inspection.
b. Findings
No findings were identified.
1EP6 Drill Evaluation
a. Inspection Scope
The inspectors observed a simulator training evolution for licensed operators on January 9, which required emergency plan implementation by an operations crew. Entergy planned for this evolution to be evaluated and included in performance indicator (PI)data regarding drill and exercise performance. The inspectors observed event classification and notification activities performed by the crew. The inspectors also attended the post-evolution critique for the scenario. The focus of the inspectors activities was to note any weaknesses and deficiencies in the crews performance and ensure that Entergy evaluators noted the same issues and entered them into the corrective action program.
b. Findings
No findings were identified.
RADIATION SAFETY
Cornerstone: Occupational/Public Radiation Safety (PS)
2RS1 Radiological Hazard Assessment and Exposure Controls
During the week of March 11 to 15, the inspectors reviewed and assessed Entergys performance in assessing the radiological hazards in the workplace associated with the implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures. The inspectors verified that Entergy is properly identifying and reporting performance indicators (PIs) for the Occupational Radiation Safety Cornerstone and identifying those performance deficiencies that were reportable as a PI and which may have represented a substantial potential for overexposure of the worker. The inspectors used the requirements in 10 CFR 20, Standards for Protection Against Radiation, Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas for Nuclear Plants, the TS, and Entergys procedures as criteria for determining compliance.
a. Inspection Scope
Radiological Hazard Assessment The inspectors selected radiologically risk-significant work activities that involved exposure to radiation being performed during RFO 30. The inspectors verified that appropriate pre-work surveys were performed to identify and quantify the radiological hazard and to establish adequate protective measures. The inspectors evaluated the radiological survey program to determine if hazards were properly identified, including the following:
identification of hot particles presence of alpha emitters potential for airborne radioactive materials, including the potential presence of transuranics and/or other hard-to-detect radioactive materials hazards associated with work activities that could suddenly and severely increase radiological conditions severe radiation field dose gradients that can result in non-uniform exposures of the body Instructions to Workers The inspectors reviewed radiation work permits (RWPs) used to access high radiation areas (HRAs) and identify what work control instructions or control barriers had been specified. The inspectors verified that allowable stay times or permissible dose for radiologically significant work under each RWP was clearly identified. The inspectors verified that electronic personal dosimeter (EPD) alarm set points were in conformance with survey indications and plant policy.
Radiological Hazards Control and Work Coverage During tours of the facility and review of ongoing work the inspectors evaluated ambient radiological conditions. The inspectors verified that existing conditions were consistent with posted surveys, RWPs, and worker briefings, as applicable.
During job performance observations, the inspectors verified the adequacy of radiological controls, such as required surveys, radiation protection job coverage, and contamination controls. The inspectors evaluated Entergys means of using EPDs in high noise areas as HRA monitoring devices.
The inspectors verified that radiation monitoring devices were placed on the individuals body consistent with the method that Entergy was employing to monitor dose from external radiation sources. The inspectors verified that the dosimeter was placed in the location of highest expected dose or that Entergy was properly employing an NRC-approved method of determining effective dose equivalent.
For high-radiation work areas with significant dose rate gradients (a factor of 5 or more),the inspectors reviewed the application of dosimetry to effectively monitor exposure to personnel. The inspectors verified that Entergys controls were adequate.
The inspectors reviewed RWPs for work within airborne radioactivity areas with the potential for individual worker internal exposures. The inspectors evaluated airborne radioactive controls and monitoring, including potential for significant airborne contamination. For these selected airborne radioactive material areas, the inspectors verified barrier integrity and temporary high-efficiency particulate air ventilation system operation.
Radiation Worker Performance During job performance observations, the inspectors observed radiation worker performance with respect to stated radiation protection work requirements. The inspectors determined that workers were aware of the significant radiological conditions in their workplace and the RWP controls/limits in place and that their performance reflected the level of radiological hazards present.
The inspectors reviewed radiological problem reports since the last inspection that found the cause of the event to be human performance errors. The inspectors determined that there was no observable pattern traceable to a similar cause. The inspectors determined that this perspective matched the corrective action approach taken by Entergy to resolve the reported problems. The inspectors discussed with the Radiation Protection Manager any problems with the corrective actions planned or taken.
Radiation Protection Technician Proficiency During job performance observations, the inspectors observed the performance of the radiation protection technician with respect to radiation protection work requirements.
The inspectors determined that technicians were aware of the radiological conditions in their workplace and the RWP controls/limits and that their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.
The inspectors reviewed radiological problem reports since the last inspection that found the cause of the event to be radiation protection technician error. The inspectors determined that there was no observable pattern traceable to a similar cause. The inspectors determined that this perspective matched the corrective action approach taken by Entergy to resolve the reported problems.
b. Findings
No findings were identified.
2RS2 Occupational As Low As is Reasonably Achievable Planning and Controls
During the week of March 11 to 15, the inspectors assessed performance with respect to maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspectors used the requirements in 10 CFR 20, Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Plants will be As Low As Reasonably Achievable, Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation Exposure As Low as Reasonably Achievable, TS, and Entergys procedures as criteria for determining compliance.
a. Inspection Scope
Radiological Work Planning The inspectors obtained from Entergy a list of work activities ranked by actual or estimated exposure that were in progress, and selected work activities of the highest exposure significance.
The inspectors reviewed the ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements. The inspectors determined that Entergy had reasonably grouped the radiological work into work activities, based on historical precedence, industry norms, and/or special circumstances.
The inspectors verified that Entergys planning identified appropriate dose mitigation features; considered, commensurate with the risk of the work activity, alternate mitigation features; and defined reasonable dose goals. The inspectors verified that Entergys ALARA assessment had taken into account decreased worker efficiency from use of respiratory protective devices and/or heat stress mitigation equipment. The inspectors determined that Entergys work planning considered the use of remote technologies as a means to reduce dose and the use of dose reduction insights from industry operating experience and plant-specific lessons learned. The inspectors verified the integration of ALARA requirements into work procedure and RWP documents.
Verification of Dose Estimates and Exposure Tracking Systems The inspectors verified that, for the selected work activities, Entergy had established measures to track, trend, and if necessary to reduce, occupational doses for ongoing work activities. The inspectors verified that dose threshold criteria were established to prompt additional reviews and/or additional ALARA planning and controls.
The inspectors evaluated Entergys method of adjusting exposure estimates, or re-planning work, when unexpected changes in scope or emergent work were encountered.
The inspectors evaluated whether adjustments to exposure estimates were based on sound radiation protection and ALARA principles and whether the frequency of adjustments supported the adequacy of the original ALARA planning process.
Radiation Worker Performance The inspectors observed radiation worker and radiation protection technician performance during work activities being performed in radiation areas, airborne radioactivity areas, or high radiation areas. The inspectors concentrated on work activities that present the greatest radiological risk to workers. The inspectors determined that workers demonstrated the ALARA philosophy in practice and that there were no procedure compliance issues. Also, the inspectors observed radiation worker performance to determine whether the training and skill level was sufficient with respect to the radiological hazards and the work involved.
b. Findings
No findings were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification
Initiating Events Cornerstone ===
a. Inspection Scope
The inspectors reviewed Entergys submittals and PI data for the indicators listed below for the period from January 2012 to December 2012. To determine the accuracy of the performance indicator data reported during those periods, the inspectors used definitions and guidance contained in the Nuclear Energy Institute (NEI) Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6. The inspectors reviewed Entergys operator narrative logs, operability assessments, maintenance rule records, condition reports, event reports, and NRC integrated inspection reports to validate the accuracy of the submittals.
Unplanned scrams per 7000 critical hours Unplanned power changes per 7000 critical hours Unplanned scrams with complications
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution
.1 Routine Review of Problem Identification and Resolution Activities
a. Inspection Scope
As required by Inspection Procedure 71152, Problem Identification and Resolution, the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that Entergy entered issues into their corrective action program at an appropriate threshold, gave adequate attention to timely corrective actions, and identified and addressed adverse trends. In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the corrective action program and periodically attended condition report review group meetings.
b. Findings
No findings were identified.
.2 Annual Sample: Review of the Operator Workaround Program
a. Inspection Scope
The inspectors reviewed the cumulative effects of the existing operator workarounds, operator burdens, existing operator aids and disabled alarms, and open main control room deficiencies to identify any effect on emergency operating procedure operator actions, and any impact on possible initiating events and mitigating systems. The inspectors evaluated whether station personnel had identified, assessed, and reviewed operator workarounds as specified in EN-FAP-OP-0006, Operator Aggregate Impact Index Performance Indicator, Revision 0, and EN-OP-117, Operations Assessments, Revision 5.
The inspectors reviewed Entergys process to identify, prioritize and resolve main control room distractions to minimize operator burdens. The inspectors reviewed the system used to track these operator workarounds and recent Entergy self assessments of the program. The inspectors also toured the control room and discussed the current operator workarounds with the operators to ensure the items were being addressed on a schedule consistent with their relative safety significance.
b. Findings and Observations
No findings were identified.
The inspectors determined that the issues reviewed did not adversely affect the capability of the operators to implement abnormal or emergency operating procedures.
The inspectors also verified that Entergy entered operator workarounds and burdens into the corrective action program at an appropriate threshold and planned or implemented corrective actions commensurate with their safety significance. The inspectors noted that Entergy performed an operator aggregate assessment of plant deficiencies due to an index value greater than the threshold value specified by procedure. The largest number of deficiencies was associated with control room annunciators. The inspectors noted that the control room annunciators are scheduled for replacement and two annunciator panel sections were replaced during RFO 30.
.3 Annual Sample: B-UPS-1A Cell 61 Low Individual Cell Voltage and Specific Gravity
a. Inspection Scope
The inspectors performed an in-depth review of Entergys apparent cause analysis and corrective actions associated with the issue of the LPCI battery UPS-1A cell 61 being found with both low ICV and low specific gravity such that it was not operable.
Specifically, Entergy identified multiple degraded cells with either low specific gravity or low ICV during the quarterly battery surveillance on October 9, 2012, and implemented an equalizing charge, single cell charging, and weekly specific gravity and ICV checks for the cells with low ICV. Subsequently, on December 9, Entergy identified cell 61 had a specific gravity of 1.176 and an ICV of 2.02 V, which rendered battery UPS-1A inoperable.
The inspectors assessed Entergys problem identification threshold, apparent cause analysis, extent of condition reviews, compensatory actions, and the prioritization and timeliness of Entergys corrective actions to determine whether Entergy was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of Entergys corrective action program and 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. In addition, the inspectors performed field walkdowns and interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.
b. Findings and Observations
Introduction.
The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not promptly correct an adverse condition resulting in the failure of the B-UPS-1A LPCI UPS battery. Specifically, Entergy personnel did not promptly replace a degraded battery cell prior to its exceeding operability limits.
Description.
On October 9, 2012, during the performance of the quarterly surveillance on battery B-UPS-1A, Entergy personnel identified multiple degraded cells with ICVs less than 2.13 volts (V). UPS-1A and UPS-1B are normally operating 480 volt rotating uninterruptible power supplies consisting of a battery (B-UPS-1A and B-UPS-1B) and a motor generator unit that provide the safety related (i.e., on a loss of offsite power)alternating current power to motor operated LPCI valves and recirculation pump suction and discharge valves.
In particular, the ICV of B-UPS-1A cell 61 was 2.06 V. Entergy personnel measured the specific gravity of cell 61 to be 1.199. VYs TS bases state that a cell will be considered out of service if its float voltage is below 2.13 V and the specific gravity is below 1.190 at 77 degrees Fahrenheit (F). Since the specific gravity measured for each cell was greater than 1.190, Entergy determined that battery B-UPS-1A was operable, the cells with ICV less than 2.13 V were degraded, and weekly monitoring was required in order to ensure continued operability.
Entergy issued an operational decision-making issue (ODMI) with specified compensatory measures, trigger points, and actions. The ODMI assumed that any further cell degradation would not consist of step changes, being a slowly developing process based on past monitoring history. Of primary importance, the ODMI required weekly readings and prompt replacement of cells with a specific gravity less than or equal to 1.192 at 77 degrees F or an ICV less than or equal to 1.90 V. Cell 61 had a temperature corrected specific gravity of no less than 1.196 and an ICV of no less than 2.05 V through December 3. On December 9, Entergy personnel measured cell 61 and determined the temperature corrected specific gravity to be 1.176 and ICV to be 2.02 V.
Entergy personnel declared UPS-1A and A LPCI inoperable, replaced cell 61, and restored the system to operable on December 10.
Entergy personnel performed an apparent cause and determined that the ODMI-directed testing was not sufficiently frequent in order to identify further degradation prior to inoperability due to the weekly monitoring being based on prior battery history, whereas the current cell was now extraordinarily aged rather than healthy. In addition, Entergy personnel concluded that twice-a-week testing would have detected the degradation prior to exceeding the TS limit.
The inspectors reviewed Regulatory Guide 1.129, Maintenance, Testing, and Replacement of Vented Lead-Acid Storage Batteries for Nuclear Power Plants, NUREG-1433, Standard Technical Specifications General Electric Plants, BWR/4, and IEEE-450, IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications. IEEE-450 states that trending individual cell voltage does not provide indication of future battery health or performance and is not recommended. In addition, prolonged operation of cells below the ICV required to maintain the cell in a fully charged state can reduce the life expectancy of cells, and cells with ICVs less than manufacturers minimum requirements that are not recovered by single cell charging will require replacement. The manufacturers specified acceptable range for float voltage is 2.17 V to 2.26 V. In addition, IEEE-450 describes an ICV of 2.07 V or below (for typical 1.215 specific gravity cells and which is not otherwise due to environmental conditions) as indicating internal cell problems that may require cell replacement. Therefore, the inspectors determined that neither weekly monitoring nor biweekly monitoring was appropriate for cells remaining below 2.07 V following single cell charging. The inspectors concluded that these results indicated extraordinarily aged cells that required prompt replacement and could not be reliably trended in order to ensure continued operability. The inspectors determined this finding was NRC identified rather than self-revealing due to Entergys apparent cause evaluation that concluded biweekly monitoring and continued trending was acceptable.
Entergys corrective actions included replacing cell 61, replacing all cells with ICVs less than 2.13 V, expediting complete battery bank replacements with a due date of May 30, and initiating CR-VTY-2012-06018.
Analysis.
The inspectors determined that Entergy personnels decision not to replace battery UPS-1A cell 61 prior to its exceeding a predetermined operability limit was a performance deficiency that was reasonably within Entergys ability to foresee and correct and should have been prevented. This finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, UPS-1A had unplanned inoperability and degraded capacity due to cell 61 being out of service which commenced at some unknown point between December 3 and December 9 and was restored when cell 61 was replaced on December 10.
In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because it did not represent a loss of system safety function or a loss of safety function for a single train (UPS-1A and A LPCI) for greater than its TS allowed outage time (seven days).
The inspectors determined that the finding had a cross-cutting aspect in the Problem Identification and Resolution area, Operating Experience component, because Entergy personnel did not implement and institutionalize available operating experience by not utilizing industry guidance contained within IEEE-450, or alternatively, vendor recommendations, to support plant safety P.2(b).
Enforcement.
10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above, from December 3, 2012, to December 10, 2012, Entergy failed to promptly correct the deficient battery UPS-1A cell 61. Entergys corrective action to restore compliance consisted of replacing battery UPS-1A cell 61 on December 10. Because this violation was of very low safety significance (Green), and Entergy entered this issue into their corrective action program (CR-VTY-2012-06018), this violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. (NCV 05000271/2013002-03, Inadequate Corrective Action for Maintaining Operability of the Low Pressure Coolant Injection Battery UPS-1A)
4OA3 Event Follow-up
a. Inspection scope
On March 18, in the process of restoring the reactor building non-safety related ventilation system to service, Entergy attempted to start the A train of the ventilation system. As expected, both the A train intake damper and fan started to ventilate and draw air into the building. However, the exhaust damper of the A train did not open due to on-going electrical maintenance at the time. The failure of the exhaust damper to open resulted in an increase of air pressure within the reactor building (i.e., secondary containment). This increase in pressure caused one of the refueling floor blow-out panels on the west side of the building to release as designed. When this occurred, operators quickly recognized the failure of the exhaust damper to open, secured the A train, and re-started the B train to re-establish a negative pressure within secondary containment in less than two minutes. The negative building pressure stopped any airflow release from the refueling floor. Entergy declared secondary containment inoperable due to the dislodged panel. NRC review determined that, at that time, secondary containment was not required to be in-service due to no fuel movement, no core alterations, and no operations with potential to drain the reactor vessel in progress.
Further, no activities generating airborne radioactive material were occurring on the refueling floor during the event.
Entergy performed, for purposes of reportability and dose assessment, bounding calculations to determine the potential amount of radioactivity released during the short period of time that secondary containment was not intact along with positive pressure within the building. Entergy determined no reportability criteria had been tripped by the short duration release, nor had any significant occupational or public dose been sustained. Entergy determined the conservative bounding projected dose at limiting sector was less than 0.0003% of the quarterly dose limit and less than 0.0002% of the annual limit. Further, Energy determined no Emergency Action Level declaration criteria had been reached. The inspectors performed an in-office and field review of Entergys calculations, non-reportability bases, non-event declaration bases, compensatory measures while the opening existed, and corrective actions restoring operability of secondary containment.
b. Findings
No findings were identified.
4OA5 Other Activities
The 4th quarter 2012 integrated inspection report, 05000271/2012005, documented an apparent violation (AV), AV 05000271/2012005-01, Failure of the B Emergency Diesel Generator from Jacket Water Leakage Due to Inadequate Corrective Action. The inspectors and an NRC SRA completed the significance determination which allowed closure of the AV to an NCV as discussed below.
Introduction.
A self-revealing NCV of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified because Entergy did not promptly correct an adverse condition resulting in the failure of the B EDG. Specifically, Entergy personnel did not promptly replace a degraded jacket water flange gasket prior to its subsequent failure.
Description.
On April 16, 2012, Entergy personnel identified a small jacket water leak on the B EDG during a monthly surveillance run. Once the diesel was running, Entergy personnel identified six drops per minute leaking from the number five opposite control side jacket water outlet jumper to header flanged gasket connection and initiated CR-VTY-2012-01772. The leak stopped after 30 minutes of operation, and the diesel was operated for about three hours.
Jacket water cools the diesel engine by circulating jacket water in a closed system with a capacity that is maintained via an expansion tank. The expansion tank is maintained at least one-half full, representing over 28 gallons of additional jacket water available for the system. However, during a design basis event, the demineralized water system does not provide additional makeup water to the expansion tank. Entergy staff declared the B EDG operable based on the small rate of the leak relative to the available jacket water contained within the expansion tank. Entergy closed CR-VTY-2012-01772 to the work management process stating that the issue did not represent an adverse condition that is required to be corrected within the corrective action process. Entergy personnel monitored the leak during subsequent monthly surveillances, and the characterization remained unchanged.
On October 15, during a monthly surveillance run that started at 9:36 am, the leak commenced upon diesel start as usual, but appeared to be larger and variable in rate during the initial loading process and did not stop once the system was heated.
Subsequently, the gasket failed, resulting in a steady, pressurized stream of jacket water. The operators promptly unloaded and secured the B EDG at 10:05 am. The inspectors interviewed auxiliary operators who estimated the final leak rate at one gallon of water every ten minutes. After replacing the gasket, Entergy restored the B EDG to operable status on October 16 at 6:55 pm.
The inspectors reviewed the apparent cause evaluation and concluded the gasket connection had failed once the jacket water system cooled during the last successful surveillance on September 10, representing 37 days of unavailability. The inspectors did not find specific operating experience for sudden failures of these gasketed connections.
However, the inspectors concluded that sudden failure of a leaking gasketed connection that was not designed or expected to leak is a generally reasonable outcome to foresee given sufficient time and/or system perturbations, and the time from April 16 to September 10 exceeded a reasonable time for prompt corrective action.
The inspectors determined that the original leak was a condition that could credibly impact nuclear safety, and therefore was a condition adverse to quality, in accordance with EN-LI-102, Corrective Action Process. EN-LI-102 requires conditions adverse to quality to be addressed in a manner that ensures timely correction of the originally identified condition. The inspectors also noted that EN-OP-104, Operability Determination Process, provides permissible classifications for operability with the closest example referring to oil leakage, a closed system with a limited reservoir capacity, from safety-related equipment that is assumed to require extended operation per the UFSAR. The inspectors determined that Entergys operability classification of Operable rather than Operable-Op Eval was not in accordance with EN-OP-104 and Entergys closure of the condition report citing no adverse condition and a work request was not in accordance with EN-LI-102.
Entergys corrective actions included replacing the gasket, visually inspecting the other jacket water connections, confirming no other similar leaks were present on either the A or B EDG, and initiating CR-VTY-2012-05044.
Analysis.
The inspectors determined that Entergy personnels decision not to repair the leaking jacket water outlet jumper to header flanged gasket connection prior to its failure in service was a performance deficiency that was reasonably within Entergys ability to foresee and correct and should have been prevented. The finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the B EDG failed in service due to a known degraded condition that affected the overall system redundancy and reliability and resulted in 37 days of unplanned unavailability.
Because the issue resulted in an actual loss of function of the B EDG for longer than its TS allowed outage time, the inspectors and a Region I SRA completed a DRE in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined this finding to be of very low safety significance (Green). The DRE estimated the increase in CDF and the increase in the frequency of a LERF as a result of the finding.
The DRE was completed using the VY standardized plant analysis risk (SPAR) model version 8.19, dated May 2012 and the following assumptions:
The condition existed for an exposure period of 37 days, given NRC determination that the jacket water leak would have likely developed the next time the B EDG started after the last successfully completed surveillance test.
The B EDG likely would have operated given the jacket water leakage for at least two hours, without any operator action to refill the jacket water expansion tank.
The condition resulted in an increased common cause failure probability for the A EDG.
The following possible responses to extend B EDG operation past two hours were not credited for conservatism:
Actions taken by operators to refill the jacket water expansion tank from the installed, but not proceduralized, service water fill line.
Actions taken by the operators/maintenance personnel to tighten the bolted connection to lessen leakage.
Recovery of the EDG if it was shutdown and the gasket quickly replaced.
The SRA also updated the VY SPAR model to include:
Credit for the two hour run time, which provided the operators an improved chance of recovering offsite power or aligning the Vernon Dam station blackout power supply.
Consequential losses of offsite power as initiating events, which increased the chance that the EDGs would be needed.
The DRE estimated a CDF for internal initiating events as a result of the finding in the range of 1 core damage accident in 2,000,000 years of reactor operation, in the mid-E-7 range per year. Grid related loss of offsite power dominated, with failure of the B EDG after two hours and the subsequent failure of the RCIC system, the high pressure coolant injection (HPCI) system and the automatic depressurization system (ADS)because of the discharge of both safety-related direct current (DC) batteries. Failure of a 480 V alternating current (AC) bus would remove AC power from the battery charger supplying DC power to RCIC and ADS and also prevent the cross tie of AC power from an installed small backup diesel generator to power a battery charger for supplying DC power to HPCI and ADS.
In accordance with IMC 0609, Appendix A, given a finding with an internal events CDF above 1E-7, the SRA used the revised SPAR model and information from the VY Individual Plant Examination, the VY Individual Plant Examination for External Events, and Entergys current probabilistic risk assessment model to estimate the impact of the finding on total CDF and LERF. Considering external initiating events such as fire, seismic and flooding, the total CDF for these would not have increased above 1 E-6 per year. This review included consideration of fire and internal flooding sequences that could cause both a loss of offsite power and loss of the A EDG. The LERF associated with the internal event CDF core damage sequences was not above 1E-7 per year. The assumptions for the LERF analysis were that all CDF station black-out sequences which resulted in core damage at high pressure resulted in a large early release and there was a CDF to LERF conversion factor of 0.1 for CDF sequences that retained that ability to remove decay heat from the containment.
The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Decision-Making, because Entergy personnel did not use conservative assumptions in decision making in that the chosen action was to monitor the leak for a prolonged period of time instead of replacing the gasket H.1(b).
Enforcement.
10 CFR 50 Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected. Contrary to the above, from September 10, 2012, to October 15, 2012, Entergy failed to promptly correct the deficient number five opposite control side jacket water outlet jumper to header flanged gasket connection. Entergys corrective action to restore compliance consisted of replacing the gasketed connection on October 15. Because this violation was of very low safety significance (Green), and Entergy entered the issue into the corrective action program (CR-VTY-2012-05044), this violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. (NCV 05000271/2013002-04, Failure of the B Emergency Diesel Generator from Jacket Water Leakage Due to Inadequate Corrective Action)
4OA6 Meetings, Including Exit
On March 15, the inspectors presented the radiation safety inspection results to Mr.
Christopher Wamser, Site Vice President, and other members of the Entergy staff. The inspectors verified that no proprietary information was retained by the inspectors or documented in this report.
On April 15, the inspectors presented the inspection results to Mr. Vincent Fallacara, General Manager of Plant Operations, and other members of the Entergy staff. The inspectors verified that no proprietary information was retained by the inspectors or documented in this report.
ATTACHMENT:
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Vermont Yankee Personnel
- C. Wamser, Site Vice President
- V. Fallacara, General Manager of Plant Operations
- M. Romeo, Director of Nuclear Safety
- J. Boyle, Engineering Director
- J. Bengtson, CA&A Manager
- R. Busick, Asst. Operations Manager
- T. Capelletti, Mechanical Superintendent
- G. Cogdon, Control Room Supervisor
- P. Corbett, Quality Assurance Manager
- D. Deer, Control Room Supervisor
- V. Ferrizzi, Shift Manager
- S. Goodman, Mechanical Maintenance Supervisor
- J. Hardy, Chemistry Manager
- E. Harms, Asst. Operations Manager
- R. Heathwaite, Chemistry Supervisor
- D. Jones, Operations Manager
- M. McKenney, Emergency Preparedness Manager
- P. McKenney, Material, Purchasing and Contracts Manager
- J. Mully, System Engineer
- J. Rogers, Design Engineering Manager
- P. Ryan, Security Manager
- K. Stupak, Manager, Training and Development
- K. Sweet, Programs and Components Engineering Supervisor
- J. Taylor, Operations Training Superintendent
- D. Tkatch, Radiation Protection Manager
- R. Wanczyk, Licensing Manager
- K. Whippie, Chemistry Supervisor
- A. Zander, Shift Manager
LIST OF ITEMS OPENED, CLOSED, DISCUSSED AND UPDATED
Opened/Closed
- 05000271/2013002-01 NCV Appendix R Fire Door Not Latching Closed Due to Misalignment (Section 1R04)
- 05000271/2013002-02 NCV Failure to Implement Compensatory Measures Associated with a Temporary Modification (Section 1R18)
- 05000271/2013002-03 NCV Inadequate Corrective Action for Maintaining Operability of the Low Pressure Coolant Injection Battery UPS-1A (Section 4OA2)
- 05000271/2013002-04 NCV Failure of the B Emergency Diesel Generator from Jacket Water Leakage Due to Inadequate Corrective Action (Section 4OA5)
Closed
- 05000271/2012005-01 AV Failure of the B Emergency Diesel Generator from Jacket Water Leakage Due to Inadequate Corrective Action(Section 4OA5)