IR 05000395/2013004
| ML13316B284 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 11/12/2013 |
| From: | Mark King NRC/RGN-II/DRP/RPB5 |
| To: | Gatlin T South Carolina Electric & Gas Co |
| References | |
| IR-13-004 | |
| Download: ML13316B284 (40) | |
Text
UNITED STATES ember 12, 2013
SUBJECT:
VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 - NRC INTEGRATED INSPECTION REPORT 05000395/2013004
Dear Mr. Gatlin:
On September 30, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Virgil C. Summer Nuclear Station, Unit 1. On October 29, 2013, the NRC inspectors discussed the results of this inspection with you and members of your staff.
Inspectors documented the results of this inspection in the enclosed inspection report.
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.
NRC inspectors documented two NRC-identifed findings and one self-revealing finding of very low safety significance (Green) in this report. Two of these findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the NRC Enforcement Policy. If you contest the violations or significance of the NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the United States Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington DC 20555-0001, with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Virgil C.
Summer Nuclear Station, Unit 1.
Additionally, if you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 Virgil C. Summer Nuclear Station, Unit 1. 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 NRCs Agencywide Document Access and management System (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Michael King, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket No.: 50-395 License No.: NPF-12
Enclosure:
NRC Integrated Inspection Report 05000395/2013004 w/Attachment: Supplemental Information
REGION II==
Docket No. 50-395 License No. NPF-12 Report No. 05000395/2013004 Licensee: South Carolina Electric & Gas (SCE&G) Company Facility: Virgil C. Summer Nuclear Station, Unit 1 Location: P.O. Box 88 Jenkinsville, SC 29065 Dates: July 1, 2013 through September 30, 2013 Inspectors: J. Reece, Senior Resident Inspector E. Coffman, Resident Inspector R. Kellner, Health Physicist (Sections 2RS1-S4, 4OA1.2, 4OA1.3)
R. Hamilton, Senior Health Physicist (Sections 2RS1-S4, 4OA1.2, 4OA1.3)
W. Pursley, Health Physicist (Sections 2RS1-S4, 4OA1.2, 4OA1.3)
Approved by: Michael King, Chief Reactor Projects Branch 5 Division of Reactor Projects Enclosure
SUMMARY OF FINDINGS
IR 05000395/2013004; 7/01/2013 - 9/30/2013: Virgil C. Summer Nuclear Station, Unit 1;
Radiological Hazard Assessment and Exposure Controls; Identification and Resolution of Problems and Other Activities The report covered a three month period of inspection by resident inspectors and three health physicists from the region. Two NRC-identified findings and one self-revealing finding were identified and two 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 aspects were 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: Mitigating System
- Green.
The inspectors identified a Green finding of licensee procedure, SAP-999,
Corrective Action, Revision 11 for the failure to develop corrective actions for a Level 3 condition report (CR) which described a condition adverse to quality (CAQ)associated with the loss of transformer, XTF-5052, alternate AC source.
The inspectors determined that the failure to develop corrective actions for a Level 3 CR as required by their corrective action program (CAP) procedure was a performance deficiency (PD). The inspectors reviewed inspector manual chapter (IMC) 0612 and determined the PD is more than minor and therefore a finding because it adversely impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and the respective attribute of equipment performance. Specifically, a system component, XTF-5052, associated with recovery of an offsite power circuit and installed to reduce core damage frequency, was rendered unavailable. The inspectors reviewed IMC 0609,
Attachment 4, and Appendix A - Exhibit 2, and determined the finding was of very low safety significance or Green because the finding was not a design deficiency or a loss of function during a required alignment per Technical Specifications. The inspectors reviewed IMC 0310 and determined the cause of the finding involved the cross-cutting area of problem identification and resolution, the component of corrective action program, and the aspect of appropriate corrective actions, P.1(d),
because the licensee failed to develop corrective actions for the Level 3 CR associated with the loss of XTF5052. (Section 4OA2.3)
- Green.
The NRC inspectors identified a NCV of 10 CFR Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately accomplish engineering services procedure, ES-0419, Equal To/Better Than (ETBT)
Evaluation Process, Revision 9, for new fire hoses located in the reactor building.
The issue was entered into the licensees CAP as condition report CR-12-05730.
The inspectors determined that the failure to adequately accomplish ES-0419 for new fire hoses in the reactor building was a PD. The inspectors reviewed IMC 0612 and determined that the PD is more than minor and therefore a finding because if left uncorrected it would have the potential to lead to a more significant safety concern in that degradation of the fire hoses in post loss of coolant accident (LOCA) conditions would adversely impact the emergency core cooling system (ECCS) containment sump screens. The inspectors reviewed IMC 0609, Attachment 4, and Appendix A -
Exhibit 2, and determined the finding was of very low safety significance or Green because the finding was not a design deficiency or represented a loss of function.
Specifically, the inspectors identified the problem prior to the licensee incurring actual risk exposure time. The inspectors reviewed IMC 0310 and determined the cause of the finding involved the cross-cutting area of human performance, the component of resources, and the aspect of adequate emergency equipment, H.2(d),
because the licensee failed to ensure the new fire hoses would not impact safety-related components such as ECCS sump screens during post-LOCA conditions.
(Section 4OA5.2)
Cornerstone: Occupational Radiation Safety
- Green.
A Green, self-revealing, NCV was identified for the failure to perform radiological surveys required by 10 CFR 20.1501(a) to ensure the potential radiological hazards and extent of radiation levels were understood and controlled before disassembling pressurizer spray valve PCV-444C. The issue was entered in the licensees CAP as CR-12-05132.
The inspectors determined that the failure to perform radiological surveys required by 10 CFR 20.1501(a) was a PD. The inspectors determined that the PD was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation in that the licensee failed to adequately evaluate potential radiological hazards that could be present in a work area. The finding was assessed using the Occupational Radiation Safety significance determination process and was determined to be of very low safety significance (Green) because it was not an ALARA planning issue, there was no overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised because the workers were wearing electronic dosimetry that was remotely monitored and intermittent on-site HP coverage was provided. This finding had a crosscutting aspect associated with human performance, work control, H.3(a). When licensee personnel failed to identify potential changes in expected radiological conditions and incorporate those changes into the RWP requirements prior to beginning work on valve PCV-444C, appropriate radiological work controls were not established.
(Section 2RS1)
B. Licensee-Identified Findings None
REPORT DETAILS
Summary of Plant Status
The unit began the inspection period at full Rated Thermal Power (RTP) and operated at or near full RTP for the remainder of the quarter.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R04 Equipment Alignment
.1 Partial System Walkdowns
a. Inspection Scope
The inspectors conducted four partial equipment alignment walkdowns which are listed below, to evaluate the operability of selected redundant trains or backup systems with the other train or system inoperable or out of service (OOS). Correct alignment and operating conditions were determined from the applicable portions of drawings, system operating procedures (SOP), and technical specifications (TS). The inspections included review of outstanding maintenance work orders (WO) and related condition reports (CR) to verify that the licensee had properly identified and resolved equipment alignment problems that could lead to the initiation of an event or impact mitigating system availability.
- Partial walkdown of A residual heat removal (RHR) during planned maintenance on B RHR pump and related components
- Partial walkdown of B emergency diesel generator (EDG) during unplanned maintenance on A EDG
- Partial walkdown of B motor driven emergency feedwater (MDEFW) and turbine driven emergency feedwater (TDEFW) during planned maintenance on A MDEFW pump
- Partial walkdown of B safety injection (SI) pump and related components while the A SI train was inoperable due to ten year motor preventive maintenance
b. Findings
No findings were identified.
.2 Complete System Walkdown
a. Inspection Scope
The inspectors performed a detailed review and walkdown of the 125 vital electrical distribution system to identify any discrepancies between the current operating system equipment lineup and the designed lineup. In addition, the inspectors reviewed SOPs, applicable sections of the final safety analysis report (FSAR), design basis document, plant drawings, completed surveillance procedures, outstanding WOs, system health reports, and related CRs to verify that the licensee had properly identified and resolved equipment problems that could affect the availability and operability of the system.
Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
1R05 Fire Protection
.1 Quarterly Fire Protection Walkdowns
a. Inspection Scope
The inspectors reviewed recent CRs, WOs, and impairments associated with the fire protection system. The inspectors reviewed surveillance activities to determine whether they supported the operability and availability of the fire protection system. The inspectors assessed the material condition of the active and passive fire protection systems and features, and observed the control of transient combustibles and ignition sources. The inspectors conducted routine inspections of the following five areas (respective fire zones also noted):
- Control building cable spreading rooms (fire zones CB-4 and CB-15)
- Relay room solid state protection system (SSPS) instrumentation and inverter (fire zones CB-6, CB-10 and CB-12)
- Intermediate building general areas (fire zones IB-25.1.1, 1.2, 1.3 and 1.5)
- Auxiliary building 388/397 elevations (fire zone AB-1.4)
- Auxiliary building 436 elevation (fire zone AB-1.18)
b. Findings
No findings were identified.
.2 Annual Fire Brigade Drill Observation
a. Inspection Scope
The inspectors observed the performance of an unannounced fire brigade drill on September 11, 2013. The inspectors evaluated the readiness of licensee personnel to respond and fight fires including the following aspects:
- Observe whether turnout clothing and self-contained breathing apparatus equipment were properly worn
- Determine whether fire hose lines were properly laid out and nozzle pattern simulated being tested prior to entering the fire area of concern
- Verify that the fire area was entered in a controlled manner
- Review whether sufficient firefighting equipment was brought to the scene by the fire brigade to properly perform their firefighting duties
- Verify that the fire brigade leaders fire fighting directions were thorough, clear and effective, and that, if necessary, offsite fire team assistance was requested
- Verify that radio communications with plant operators and between fire brigade members were efficient and effective
- Confirm that fire brigade members checked for fire victims and fire propagation into applicable plant areas
- Observe if effective smoke removal operations were simulated
- Verify that the fire fighting pre-plans were properly utilized and were effective
- Verify that the licensee pre-planned drill scenario was followed, drill objectives met the acceptance criteria, and deficiencies were captured in post drill critiques
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
.1 Resident Quarterly Review of Operator Requalification
a. Inspection Scope
The inspectors observed an operator requalification scenario occurring on September 16, 2013. The scenario involved the following failures/events: a main turbine first stage steam pressure channel, a main feedwater pump, a steam generator tube leak, a steam generator feedwater control valve and a loss of offsite power. 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 and when required, emergency action levels as the Site Emergency Director. The inspectors reviewed the licensees critique comments to verify that performance deficiencies were captured for appropriate corrective action.
b. Findings
No findings were identified.
.2 Resident Quarterly Observation of Control Room Operations
a. Inspection Scope
During the inspection period, the inspectors conducted observations of licensed reactor operator activities to ensure consistency with licensee procedures and regulatory requirements. For the two listed activities, the inspectors observed the following elements of operator performance: 1) operator compliance and use of plant procedures including technical specifications; 2) control board component manipulations; 3) use and interpretation of plant instrumentation and alarms; 4) documentation of activities; 5) management and supervision of activities; and 6) control room communications.
- Observation of pressurizer level and pressure surveillance test procedures (STP);reactor building (RB) air addition
- Observation of power range nuclear instruments (NIs) calibration and testing
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors evaluated two equipment issues described in the CRs listed below to verify the licensees effectiveness with the corresponding preventive or corrective maintenance associated with structure, system, and components (SSCs). The inspectors reviewed Maintenance Rule (MR) implementation to verify that component and equipment failures were identified, entered, and scoped within the MR program.
Selected SSCs were reviewed to verify proper categorization and classification in accordance with 10 CFR 50.65. The inspectors examined the licensees 10 CFR 50.65(a)(1) corrective action plans to determine if the licensee was identifying issues related to the MR at an appropriate threshold and that corrective actions were established and effective. The inspectors review also evaluated if maintenance preventable functional failures or other MR findings existed that the licensee had not identified.
The inspectors reviewed the licensees controlling procedures consisting of engineering services procedure (ES)-514, Rev. 6, Maintenance Rule Program Implementation, and station administrative procedure (SAP)-0157, Rev. 1, Maintenance Rule Program, to verify consistency with the MR program requirements.
- CR-13-02406, Maintenance Rule
- (a1) goal setting established for the local ventilation cooling system for adequate ventilation of the A SI pump room
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessment and Emergent Work Control
a. Inspection Scope
The inspectors performed risk assessments, as appropriate, for the five selected work activities listed below: 1) the effectiveness of the risk assessments performed before maintenance activities were conducted; 2) the management of risk; 3) that, upon identification of an unforeseen situation, necessary steps were taken to plan and control the resulting emergent work activities; and, 4) that emergent work problems were adequately identified and resolved. The inspectors evaluated the licensees work prioritization and risk characterization to determine, as appropriate, whether necessary steps were properly planned, controlled, and executed for the planned and emergent work activities.
- Work Week 28, risk assessment for planned maintenance on B train RHR pump and related components that resulted in a yellow risk condition
- Work Week 30, risk assessment for unplanned maintenance on A EDG resulting in a yellow risk condition
- Work Week 35, risk assessment for loss offsite power times four (X4) during testing at Parr Hydro
- Work Week 39, yellow risk assessment for A train service water (SW) pump motor maintenance
- Work Week 40, yellow risk assessment for C train SW pump motor maintenance
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments
a. Inspection Scope
The inspectors reviewed six operability evaluations listed below, affecting risk significant mitigating systems to assess, as appropriate: 1) the technical adequacy of the evaluations; 2) whether operability was properly justified and the subject component or system remained available, such that no unrecognized increase in risk occurred; 3) whether other existing degraded conditions were considered; 4) that the licensee considered other degraded conditions and their impact on compensatory measures for the condition being evaluated; and, 5) the impact on TS limiting conditions for operations and the risk significance in accordance with the significance determination process. The inspectors also verified that the operability evaluations were performed in accordance with SAP-209, Rev. 1, Operability Determination Process, and SAP-999, Rev. 11, Corrective Action Program.
- CR-13-02694, A chiller tripped twice on circuit #1 pump down exceeded time limit
- CR-13-01471, SW to component cooling loop B cross connect valve failed to fully open
- CR-13-02143, perform past operability evaluation for the B RB spray system
- CR-13-02646, oil sample on XPP0021B, B MDEFW pump, indicated elevated metal contents
b. Findings
No findings were identified.
1R18 Plant Modifications
a. Inspection Scope
The inspectors reviewed one temporary modification associated with work order (WO)1308097 that involved the installation of jumpers for a tagout to support work activities associated with the auxiliary building charcoal exhaust filters to evaluate the change for adverse effects on system availability, reliability, and functional capability. Documents reviewed included engineering calculations, WOs, site drawings, applicable sections of the FSAR, supporting 10 CFR 50.59 evaluations, TS, and design basis information. The inspectors evaluated the change documents and associated 10 CFR 50.59 reviews against the system design basis documentation and FSAR to verify that the changes did not adversely affect the safety function of safety systems. The inspectors also reviewed any related CRs to confirm that problems were identified at an appropriate threshold, were entered into the corrective action program (CAP), and appropriate corrective actions had been initiated
b. Findings
No findings were identified.
1R19 Post Maintenance Testing
a. Inspection Scope
For the six maintenance activities listed below, the inspectors reviewed the associated post-maintenance testing (PMT) procedures and either witnessed the testing and/or reviewed test records 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) test 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 its safety function. The inspectors verified that these activities were performed in accordance with general test procedure (GTP)-214, Post Maintenance Testing Guideline, Rev. 5, Change B.
- WO 1304603-001, perform motor load testing on C TDEFW steam supply valve following packing adjustment
- WO 1310823-003 and 004, perform A chiller run following a change to local and main breaker settings
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed and/or reviewed the six surveillance test procedures (STPs)listed below to verify that TS or risk significant surveillance requirements were followed and that test acceptance criteria were properly specified to ensure that the equipment could perform its intended safety function. The inspectors verified that proper test conditions were established as specified in the procedures, that no equipment preconditioning activities occurred, and that acceptance criteria were met.
In-Service Tests:
- STP- 220.002, Turbine Driven Emergency Feedwater Pump and Valve Test, Rev. 7D
- STP- 205.004, RHR Pump and Valve Operability Test, Rev. 7
- STP- 220.001A, Motor Driven Emergency Feedwater Pump and Valve Test, Rev. 9A
- STP-112.003, Reactor Building Spray System Valve Operability Test, Rev. 9C
- STP-120.004, Emergency Feedwater Valve Operability Test, Rev. 16 Other:
- STP-310.007, NIS Power Range N43 Calibration, Rev. 12C
b. Findings
No findings were identified.
RADIATION SAFETY
2RS1 Radiological Hazard Assessment and Exposure Controls
a. Inspection Scope
Hazard Assessment and Instructions to workers During facility tours, the inspectors directly observed labeling of radioactive material and postings for radiation areas, high radiation areas (HRAs), and airborne radioactivity areas established within the radiologically controlled area (RCA) of the auxiliary building and radioactive waste (radwaste) processing and storage locations. The inspectors independently measured radiation dose rates or directly observed conduct of licensee radiation surveys for selected RCA areas. The inspectors reviewed survey records for several plant areas including surveys for alpha emitters, airborne radioactivity, gamma surveys with a range of dose rate gradients, and pre-job surveys for several U1RFO20 (Unit 1 Refueling Outage 20) outage tasks including reactor head repair activities. The inspectors also discussed changes to plant operations that could contribute to changing radiological conditions since the last inspection. For selected work activities, the inspectors observed informal pre-job briefings and reviewed radiation work permit (RWP) details to assess communication of radiological control requirements and current radiological conditions to workers.
Hazard Control and Work Practices The inspectors evaluated access barrier effectiveness for selected Locked High Radiation Area (LHRA) locations and discussed changes to procedural guidance for LHRA and Very High Radiation Area (VHRA)controls with health physics (HP) supervisors. The inspectors observed and evaluated controls for the storage of irradiated material within the spent fuel pool (SFP).
Established radiological controls (including airborne controls) were evaluated for selected work activities. In addition, the inspectors reviewed and discussed licensee controls for areas where dose rates could change significantly.
Through direct observations and interviews with licensee staff, the inspectors evaluated occupational workers adherence to selected RWPs and HP technician (HPT) proficiency in providing job coverage. Electronic dosimeter (ED) alarm set points and worker stay times were evaluated against area radiation survey results for selected U1RFO20 work activities. The inspectors reviewed the use of personnel dosimetry (extremity dosimetry and multibadging in high dose rate gradients) for the head repair work completed during U1RFO20. The inspectors also evaluated worker response to dose and dose rate alarms during selected work activities.
Control of Radioactive Material The inspectors observed surveys of material and personnel being released from the RCA using small article monitor, personnel contamination monitor, and portal monitor instruments. The inspectors also reviewed licensee procedures and methodology for releasing concrete rubble and soil from inside the RCA associated with construction activities for the Independent Spent Fuel Storage Facility (ISFSI). As part of Inspection Procedure (IP) 71124.05, the inspectors reviewed the calibration records for selected release point survey instruments and discussed equipment sensitivity, alarm setpoints, and release program guidance with licensee staff.
Also as part of IP 71124.05, the inspectors compared recent 10 Code of Federal Regulations (CFR) Part 61 results for the Dry Active Waste (DAW) radioactive waste stream with radionuclides used in calibration sources to evaluate the appropriateness and accuracy of release survey instrumentation. The inspectors reviewed records of leak tests on selected sealed sources and discussed nationally tracked source transactions with licensee staff.
Problem Identification and Resolution The inspectors reviewed CAP documents associated with radiological hazard assessment and exposure control. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with licensee procedures. The inspectors also reviewed recent self-assessment results.
The inspectors evaluated radiation protection activities against the requirements and guidance of FSAR Section 12; TS Sections 6.8, 6.11 and 6.12; 10 CFR Parts 19 and 20; Regulatory Guide (RG) 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants; and approved licensee procedures. Licensee programs for monitoring materials and personnel released from the RCA were evaluated against 10 CFR Part 20 and IE Circular 81-07, Control of Radioactively Contaminated Material.
Documents reviewed are listed in the Attachment.
The inspectors completed 1 sample as required by IP 71124.01 (sample size of 1).
b. Findings
Introduction.
A Green, self-revealing, non-cited violation (NCV) was identified for the failure to perform radiological surveys required by 10 CFR 20.1501(a) to ensure the potential radiological hazards and extent of radiation levels were understood and controlled before disassembling pressurizer spray valve PCV-444C. The issue was entered in the licensees corrective action program as CR-12-05132.
Description.
On November 4, 2012, two valve technicians entered the reactor building pressurizer cubicle to continue work on pressurizer spray valve PCV-444C. The technicians entered the work area using RWP 12-04300, Task 7, All Work Associated with PCV000444 B, C, D & PCV00445 B in PZR, which allowed access into radiation areas and had ED alarm setpoints of 100 millirem for dose and 200 millirem/hour (mR/hr) for dose rate. The RWP required remote ED monitoring (telemetric) and intermittent HP oversight. Approximately 1.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> after signing in on the RWP, one of the workers received an ED dose rate alarm at a maximum dose rate of 628 mR/hr. The HP rover responded to the work area and found that the technicians had disassembled the valve and removed the stem from the valve body. The HP rover stopped work, had the workers exit containment, and surveyed the area. The HP rover identified high radiation dose rates of 600 mR/hr on contact with the valve stem and 160 mR/hr at 30 cm, and transferable contamination levels of 12 millirad/hour. These unexpected conditions were subsequently posted and controlled. When exiting the Radiologically Controlled Area (RCA) both workers alarmed the personnel contamination monitors.
Subsequent surveys of the workers identified widespread low levels of contamination on their personal modesty clothing and 500 counts per minute above background on the neck and chest of both workers. The workers were decontaminated and released from the RCA. The workers received 67 millirem and 54 millirem of exposure during the entry and there was no indication of internal contamination.
RWP 12-04300, Task 7, contained no special precautions or instructions for breaching the valve and no ALARA Pre-Job Briefing was required because radiological conditions for the RWP task were not expected to have elevated radiological risk. The inspectors also noted that other tasks for RWP 12-04300 contained additional precautions and requirements including continuous HP coverage when breaching systems. The inspectors additionally noted that several other pressurizer control valves had recently been worked using RWP 12-04300 with no radiological control issues. A review of HP shift turnover logs by the inspectors identified entries on November 2 and 3, 2012, prior to the ED alarm, for 444C teardown and Breach 444C in the pressurizer indicating work on PCV-444C had been ongoing for at least a day prior to the incident.
The inspectors noted that the causal evaluation completed by the licensee for CR-12-05132 identified that disassembly of valve PCV-444C was emergent outage work and the original RWP was not written to include the change in work scope. The licensees evaluation also identified that personnel aware of the work scope change did not use available historical radiological information concerning high dose rates and contamination levels associated with valve PCV-444C.
Analysis.
The inspectors determined that the failure to perform radiological surveys required by 10 CFR 20.1501(a) was a PD. The inspectors determined that the PD was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation in that the licensee failed to adequately evaluate potential radiological hazards that could be present in a work area. The finding was assessed using the Occupational Radiation Safety SDP and was determined to be of very low safety significance (Green)because it was not an ALARA planning issue, there was no overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised because the workers were wearing electronic dosimetry that was remotely monitored and intermittent on-site HP coverage was provided. This finding had a crosscutting aspect associated with human performance, work control, H.3(a). When licensee personnel failed to recognize potential changes in expected radiological conditions and incorporate those changes into the RWP requirements prior to beginning work on valve PCV-444C, appropriate radiological work controls were not established.
Enforcement.
Title 10 CFR 20.1501(a) requires, in part, that the licensee make or cause to be made, surveys that are necessary to comply with the regulations in 10 CFR Part 20 and that are reasonable under the circumstances to evaluate the potential radiological hazards that could be present. Pursuant to 10 CFR Part 20.1003, a survey is defined, in part, as an evaluation of the radiological conditions and potential hazards incident to the production, use, and presence of radioactive material or other sources of radiation.
10 CFR 20.1902(b) requires the posting of High Radiation Areas. Contrary to the above, on November 4, 2012, the licensee failed to adequately evaluate the potential radiological hazards that could be present in a work area to assure compliance with 10 CFR 20.1902(b), in that valve technicians while disassembling pressurizer spray valve PCV-444C created a high radiation area that was not properly posted or monitored.
Because this failure to perform radiological surveys is of very low safety significance and has been entered into the licensees corrective action program (CR 12-05132), this violation is being treated as a non-cited violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000395/2013004-01, Failure to Evaluate Radiological Conditions.
2RS2 Occupational ALARA Planning and Controls
a. Inspection Scope
As Low As Reasonably Achievable (ALARA) Program Status The inspectors reviewed and discussed plant exposure history and current trends including the sites three-year rolling average (TYRA) collective exposure history for calendar year (CY) 2010 through CY 2012. Current and proposed activities to manage site collective exposure and trends regarding collective exposure were evaluated through review of previous TYRA collective exposure data and review of the licensees 5-year ALARA program implementing plan. Current ALARA program guidance and recent changes, as applicable, regarding estimating and tracking exposure were discussed and evaluated.
Radiological Work Planning The inspectors reviewed planned work activities and their collective exposure estimates for U1RFO20 (Unit 1 Refueling Outage 20) work activities including the reactor vessel head project, refueling activities, major valve maintenance and pressurizer work. For the selected tasks, the inspectors reviewed dose mitigation actions and the established dose goals. The use of remote technologies including teledosimetry and remote visual monitoring were reviewed as specified in RWP or procedural guidance. Collective dose data for selected tasks were compared with estimates and, where applicable, changes to established estimates were discussed with responsible licensee ALARA planning representatives. The inspectors reviewed previous post-job reviews conducted for the U1RFO20 tasks and reviewed items entered into the licensees CAP and referred to the ALARA Committee for evaluation.
Verification of Dose Estimates and Exposure Tracking Systems The inspectors reviewed the selected ALARA work packages and discussed assumptions with responsible planning personal regarding the bases for the current estimates. The licensees on-line RWP cumulative dose data bases used to track and trend current personal and cumulative exposure data and/or to trigger additional ALARA planning activities in accordance with current procedures were reviewed and discussed.
Source Term Reduction and Control The inspectors reviewed historical dose rate trends for shutdown chemistry, cleanup, and resultant chemistry and RP trend-point data against the recent U1RFO20 data. The inspectors reviewed the correlation of the exposure trends to the various exposure reduction initiatives taken over the years with historical data and discussed with licensee staff.
Radiation Worker Performance Inspector observations of radiation worker performance in high risk work was limited because the plant did not have a RFO in 2013. The inspectors observed briefings at the RCA control point and interviewed workers and HPTs for understanding and awareness of the sites current ALARA campaign. In interviews with HPTs, the inspectors evaluated their understanding of their role in the use of remote technologies to reduce dose including teledosimetry and remote visual monitoring. Radiation worker performance was also evaluated as part of IP 71124.01.
Problem Identification and Resolution The inspectors reviewed and discussed selected CAP documents associated with ALARA program implementation. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with licensee procedure SAP-0999, Corrective Action Program, Rev. 11. The inspectors also evaluated the scope and frequency of the licensees self-assessment program and reviewed recent assessment results.
ALARA program activities were evaluated against the requirements of UFSAR Section 12; TS Sections 6.8 Procedures and Programs, 6.11 Radiation Protection, and 6.12 High Radiation Areas; 10 CFR Part 20; and approved licensee procedures. Records reviewed are listed in the Attachment.
The inspectors completed 1 sample as required by IP 71124.02 (sample size of 1).
b. Findings
No findings were identified.
2RS3 In-Plant Airborne Radioactivity Control and Mitigation
a. Inspection Scope
Engineering Controls The inspectors reviewed the use of temporary and permanent engineering controls to mitigate airborne radioactivity during transfer canal repair activities.
The inspectors observed the use of negative pressure units (NPU)s and a temporary containment (tent) and reviewed NPU testing records. Use of containment purge to reduce airborne levels in general areas was reviewed. The inspectors evaluated the effectiveness of continuous air monitors and air samplers placed in work area breathing zones to provide indication of increasing airborne levels.
Respiratory Protection Equipment The inspectors reviewed the use of respiratory protection devices to limit the intake of radioactive material. This included review of devices used for routine tasks and devices stored for use in emergency situations. The inspectors reviewed ALARA evaluations for the use of respiratory protection devices.
Selected Self-Contained Breathing Apparatus (SCBA) units and negative pressure respirators (NPRs) staged for routine and emergency use in the Main Control Room and other locations were inspected for material condition, SCBA bottle air pressure, number of units, and number of spare masks and air bottles available. The inspectors reviewed maintenance records for selected SCBA units for the past two years and evaluated SCBA and NPR compliance with National Institute for Occupational Safety and Health certification requirements. The inspectors also reviewed records of air quality testing for supplied-air devices and SCBA bottles.
The inspectors reviewed the process for issuing respiratory protection devices to workers, including verification of training and medical qualifications. The inspectors observed practical factors training for three radworkers requalification, observed the fit testing the three workers for an NPR and SCBA mask, and interviewed radworkers on use of the devices including SCBA bottle change-out and use of corrective lens inserts. Respirator qualification records and medical fitness records were reviewed for several personnel. In addition, qualifications for individuals responsible for testing and repairing SCBA vital components were evaluated through review of training records.
Problem Identification and Resolution CRs associated with airborne radioactivity mitigation and respiratory protection were reviewed and assessed. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with procedure SAP-999 Corrective Action Program, Rev. 11. Documents reviewed are listed in the Attachment.
Licensee activities associated with the use of engineering controls and respiratory protection equipment were reviewed against 10 CFR Part 20; UFSAR Chapter 12; the guidance in Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection; and applicable licensee procedures. Documents reviewed are listed Attachment.
The inspectors completed 1 sample as required by IP 71124.03 (sample size of 1).
b. Findings
No findings were identified.
2RS4 Occupational Dose Assessment
a. Inspection Scope
External Dosimetry The inspectors reviewed the licensees National Voluntary Accreditation Program (NVLAP) certification data for accreditation for 2012-2013 and 2013-2014 for Ionizing Radiation Dosimetry. The inspectors reviewed program procedures for processing active EDs and onsite storage of Thermo Luminescent Dosimeters (TLDs). Comparisons between ED and TLD results, including correction factors, were discussed in detail. The inspectors also reviewed dosimetry occurrence reports regarding alarming dosimeters.
Internal Dosimetry Inspectors reviewed and discussed the in vivo bioassay program with the licensee. Inspectors reviewed procedures that addressed methods for determining internal or external contamination, releasing contaminated individuals, the assignment of dose, and the frequency of measurements depending on the nuclides.
Inspectors reviewed and evaluated Whole Body Counter (WBC) sensitivity, count time and libraries. The inspectors discussed assessment and disposition of unexpected dosimetry results to include workers reporting for work after working abroad having received documentable uptakes. The inspectors evaluated the licensees program for in vitro monitoring, however there were no dose assessments using this method to review.
There were no internal dose assessments for internal exposure greater than 10 millirem committed effective dose equivalent to review.
Special Dosimetric Situations The inspectors reviewed records for three currently declared pregnant workers (DPW)s and discussed guidance for monitoring and instructing DPWs. Inspectors reviewed the licensees practices for monitoring external dose in areas of expected dose rate gradients, including the use of multi-badging and extremity dosimetry. The inspectors evaluated the licensees neutron dosimetry program including instrumentation which was evaluated under procedure 71124.05.
Problem Identification and Resolution The inspectors reviewed and discussed licensee CAP documents associated with occupational dose assessment. Inspectors evaluated the licensees ability to identify and resolve the identified issues in accordance with procedure SAP-0999 Corrective Action Program, Rev. 11. The inspectors also discussed the scope of the licensees internal audit program and reviewed recent assessment results.
HP program occupational dose assessment activities were evaluated against the requirements of FSAR Section 12; TS Section 6.11; 10 CFR Parts 19 and 20; and approved licensee procedures. Records reviewed are listed in the Attachment.
The inspectors completed 1 sample as required by IP 71124.04 (sample size of 1).
b. Findings
No findings were identified.
2RS5 Radiation Monitoring Instrumentation
a. Inspection Scope
The inspectors reviewed the licensees radiation monitoring instrumentation programs to verify the accuracy and operability of radiation monitoring instruments used to monitor areas, materials, and workers to ensure a radiologically safe work environment and to detect and quantify radioactive process streams and effluent releases.
Walkdowns and Observations: During tours of the auxiliary building, SFP areas, and RCA exit point, the inspectors observed installed radiation detection equipment including the following instrument types: area radiation monitors (ARM), continuous air monitors (CAM), liquid and gaseous effluent monitors, personnel contamination monitors (PCM),small article monitors (SAM), and portal monitors (PM). The inspectors observed the physical location of the components, noted the material condition, and compared sensitivity ranges with UFSAR requirements.
Calibration and Testing Program: In addition to equipment walk-downs, the inspectors observed source checks and alarm setpoint testing of various instruments, including friskers, hand and foot monitors and airborne radioactivity monitors. For the portable instruments, the inspectors discussed the use of a high-range calibrator and discussed periodic output value testing with a radiation protection technician. The inspectors reviewed the last two calibration records and evaluated alarm setpoint values for selected ARM, PCM, PM, SAM, effluent monitors, laboratory counting systems, and WBC systems. This included a sampling of instruments used for post-accident monitoring such as containment high-range ARMs, and effluent monitor high-range noble gas and iodine channels. Radioactive sources used to calibrate selected ARMs and effluent monitors were evaluated for traceability to national standards. Calibration stickers on portable survey instruments and air samplers were noted during inspection of storage areas for ready-to-use equipment. The most recent 10 CFR Part 61 analysis for DAW was reviewed to determine if calibration and check sources are representative of the plant source term. The inspectors also reviewed countroom quality assurance records for gamma ray spectroscopy equipment.
Effectiveness and reliability of selected radiation detection instruments were reviewed against details documented in the following: 10 CFR Part 20; NUREG-0737, Clarification of TMI Action Plan Requirements; UFSAR Chapters 12; and applicable licensee procedures. Documents reviewed during the inspection are listed in section 2RS5 of the report Attachment.
Problem Identification and Resolution: The inspectors reviewed and discussed selected Corrective Action Program (CAP) documents associated with radiological instrumentation. The reviewed items included CRs, self-assessment, and quality assurance audit documents. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with procedure SAP-999 Corrective Action Program, Rev. 11. Documents reviewed are listed the Attachment.
The inspectors completed 1 sample as required by IP 71124.05 (sample size of 1).
b. Findings
No findings were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification
.1 Mitigating Systems Cornerstone
a. Inspection Scope
The inspectors verified the accuracy of the licensees PI submittals listed below for the period July 2012 through June 2013. The inspectors used the performance indicator definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02, Rev. 6, Regulatory Assessment Performance Indicator Guideline, and licensee procedure SAP-1360, Rev. 1, NRC and INPO/WANO Performance Indicators, to check the reporting of each data element. The inspectors sampled licensee event reports (LERs),operator logs, plant status reports, CRs, and performance indicator data sheets to verify that the licensee had properly reported the PI data. Also, the inspectors discussed the PI data with the licensee personnel associated with the performance indicator data collection and evaluation.
- MSPI - High Head Safety Injection System
- MSPI - Residual Heat Removal System
b. Findings
No findings were identified.
.2 Occupational Radiation Safety Cornerstone
a. Inspection Scope
The inspectors reviewed the Occupational Exposure Control Effectiveness PI results for the Occupational Radiation Safety Cornerstone from June 2012 through July 2013. For the assessment period, the inspectors reviewed ED alarm logs and selected CRs related to controls for exposure significant areas. The inspectors also reviewed licensee procedural guidance for collecting and documenting PI data. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.3 Public Radiation Safety Cornerstone
a. Inspection Scope
The inspectors reviewed the Radiological Control Effluent Release Occurrences PI results for the Public Radiation Safety Cornerstone from June 2012 through July 2013.
For the assessment period, the inspectors reviewed cumulative and projected doses to the public contained in liquid and gaseous release permits and CRs related to Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual issues.
The inspectors also reviewed licensee procedural guidance for collecting and documenting PI data. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution
.1 Review of Items Entered into the Corrective Action Program
a. Inspection Scope
As required by IP 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 either attending daily screening meetings that briefly discussed major CRs, or accessing the licensees computerized corrective action database and reviewing each CR that was initiated.
b. Findings
No findings were identified.
.2 Semi-Annual Review to Identify Trends
a. Inspection Scope
The inspectors performed a review of the licensees CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue. The review was focused on repetitive equipment issues, but also considered trends in human performance errors, the results of daily inspector corrective action item screening discussed in Section 4OA2.1 above, licensee trending efforts, and licensee human performance results. The review nominally considered the six-month period of April, 2013, through September, 2013. Documents reviewed included licensee monthly and quarterly corrective action trend reports, engineering system health reports, maintenance rule documents, department self-assessment activities, and quality assurance audit reports.
b. Findings
No findings were identified. In general, the licensee has identified trends and has addressed the trends within their CAP. However, inspectors noted that Appendix R fire doors (FRA), including some doors functioning as steam propagation barriers (SPB),control room pressure boundaries (CRB) or CO2 boundaries (CO2), continue to have issues. This trend was previously discussed in inspection report 05000395/2012005 and 05000395/2011005, and the licensee continues to monitor the trend under CR-13-04356. Specifically from March 2013 to September 2013, the inspectors identified the following CRs:
Condition Description Door Type Report CR-13-01079 Door will not close under its own power SPB, FRA CR-13-01194 Door latching intermittently CO2, FRA CR-13-01424 Door will not latch FRA CR-13-02083 Lower latch broken, degraded SPB function SPB, FRA CR-13-02808 Door did not latch following use CRB CR-13-02848 Door had problems closing and latching SPB, FRA CR-13-03639 Door will not close under its own power SPB, FRA CR-13-03998 Door will not close under its own power SPB, FRA
.3 Annual Sample Review of CR-13-01962
a. Inspection Scope
The inspectors reviewed CR-13-01962, during switching at Parr substation a voltage spike occurred on low side of XTF5052 transformer, dated May 3, 2013, in detail to evaluate the effectiveness of the licensees corrective actions for important safety issues. The inspectors assessed whether the issue was properly identified, documented accurately and completely, properly classified and prioritized, adequately considered extent of condition, generic implications, common cause, and previous occurrences, adequately identified root causes/apparent causes, and identified appropriate and timely corrective actions. Also, the inspectors verified the issues were processed in accordance with procedure, SAP-999, Corrective Action Program, Rev. 11.
b. Findings
Introduction:
The inspectors identified a Green finding of licensee procedure, SAP-999, Corrective Action, Revision 11 for the failure to develop corrective actions for a Level 3 condition report (CR) which described a condition adverse to quality (CAQ) associated with the loss of transformer, XTF-5052, alternate AC source.
Description:
On May 3, 2013, the licensee initiated CR-13-01962 based on the loss of the alternate AC (AAC) circuit and respective 13.2/7.2kV transformer, XTF5052 that occurred during switching operations at the Parr switchyard. This also resulted in an unplanned Yellow risk condition as evaluated by the licensees maintenance rule risk program. Following completion of the licensees evaluation of the event and close-out of CR-13-01962, the inspectors performed a detailed review and identified the following:
- The CR was assigned a Level 3 category.
- The immediate action was identified as entering AAC transformer into their risk program because it had tripped off line and was unavailable.
- The cause evaluation stated that the probable cause was alignment to construction power that had higher voltage than the normal input to the Parr Station resulting in a breaker trip and loss of power to XTF5052.
- The proposed corrective actions section stated, XTF5052 was taken out of service as a precautionary measure until Parr was returned to normal voltage. No further actions required. The inspectors obtained additional information which indicated that after XTF-5052 had lost power it was added to the licensees Removal and Restoration program.
The inspectors reviewed the licensees CAP procedure, SAP-999, Corrective Action, Revision 11, and noted:
- Step 4.1.16 that a CAQ is defined as those conditions that have a negative impact on the performance of facility structures, systems, and components or programs which are described in the Current Licensing Basis (CLB).
- Enclosure A for CR Categorization Guidelines states, in part, for a Level 3 CR that corrective actions are developed to maintain station safety and reliability for equipment.
The inspectors reviewed the licensees nuclear licensing (NL) procedure, 123, Nuclear Licensing Management of the Current License Basis and Licensing Support Documents, Revision 1, for CLB documents and noted that Enclosure A, Current License Basis, includes the FSAR. The inspectors reviewed FSAR Section 8.2.1.1 and noted the description of the VC Summer AAC included the 13.2/7.2kV transformer.
The inspectors concluded that the Level 3 CR described a CAQ which was the alignment to construction power that had higher voltage than the normal input to the Parr Station resulting in a breaker trip and loss of power to XTF5052 or a negative impact on the performance of XTF5052. The inspectors also concluded that the licensee failed to meet the standards established by their CAP procedure to develop corrective action for a Level 3 CR because no corrective actions were developed to address the CAQ involving the alignment to construction power resulting in the loss of XTF-5052.
Analysis:
The inspectors determined that the failure to develop corrective action for a Level 3 CR as required by their CAP procedure was a PD. The inspectors reviewed inspector manual chapter (IMC) 0612 and determined the PD is more than minor and therefore a finding because it adversely impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and the respective attribute of equipment performance. Specifically, a system component, XTF-5052, associated with recovery of an offsite power circuit and installed to reduce core damage frequency, was rendered unavailable. The inspectors reviewed IMC 0609, Attachment 4, and Appendix A - Exhibit 2 for the Mitigating Systems cornerstone, and determined the finding was of very low safety significance or Green because the finding was not a design deficiency or loss of function during a required alignment per Technical Specifications. The inspectors reviewed IMC 0310 and determined the cause of the finding involved the cross-cutting area of problem identification and resolution, the component of corrective action program, and the aspect of appropriate corrective actions, P.1(d), because the licensee failed to develop corrective actions for the Level 3 CR associated with the loss of XTF5052.
Enforcement:
Licensee procedure, SAP-999, requires in part that corrective actions are developed for adverse conditions to maintain station safety and reliability for equipment.
Contrary to this, on June 4, 2013, the licensee failed to adequately develop corrective actions for a Level 3 CR involving the loss of XTF5052. Because this finding does not involve a violation of regulatory requirements, has very low safety significance, and has been entered into the licensees CAP as CR-13-03178, this finding is identified as a FIN 05000395/2013004-02, Failure to Develop Adequate Corrective Action for a Condition Adverse to Quality Involving the Loss of the Alternate AC Transformer.
4OA3 Followup of Events and Notices of Enforcement Discretion
(Closed) LER 05000395/2012-002-01: Seismically Qualified Refueling Water Storage Tank (RWST) Aligned to Non-Seismic Piping On July 31, 2013, the licensee issued LER 2012-002-01, which was revision 1 to the previous LER 2012-002-00 reviewed and closed to a NRC identified Green NCV in integrated inspection report (IR) 05000395/2012005. The inspectors reviewed the additional information added in 2012-002-01 relating to extent of condition reviews and the return paths of the purification loop back to the RWST. The inspectors identified no other findings related to the event. 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 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.
.2 (Closed) Unresolved Item (URI)05000395/2012005-02, Post-LOCA impact of
Replacement Fire Hoses on the ECCS Sump
a. Inspection Scope
The inspectors opened the above URI in NRC integrated inspection report 05000395/2012005 to allow further review of the identified PD to determine if the significance was more than minor. The inspectors have completed their review of the aforementioned URI which is hereby closed as discussed below.
b. Findings
Introduction:
The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately accomplish engineering services procedure, ES-0419, Equal To/Better Than (ETBT) Evaluation Process, Revision 9, for new fire hoses located in the reactor building.
Description:
The details of URI 05000395/2012005-02 are discussed in the above report. The inspectors reviewed the licensees evaluation which concluded that detailed evaluation of the new fire hoses for impact on emergency core cooling system (ECCS)containment sump screens would require analysis by a vendor. Because the inspectors identified the problem prior to unit startup and operation into the next cycle, subsequent review and regional discussion determined that no actual risk exposure had occurred.
The inspectors concluded that the licensee took adequate corrective actions by locating and installing additional fire hoses that were qualified for post loss of coolant accident (LOCA) conditions. However, the inspectors also concluded that the licensee failed to adequately accomplish ES-0419.
Analysis:
The inspectors determined that the failure to adequately accomplish ES-0419 for new fire hoses in the reactor building was a PD. The inspectors reviewed IMC 0612 and determined the PD is more than minor and therefore a finding because if left uncorrected it would have the potential to lead to a more significant safety concern in that degradation of the fire hoses in post-LOCA conditions would adversely impact the ECCS containment sump screens. The inspectors reviewed IMC 0609, Attachment 4, and Appendix A - Exhibit 2, and determined the finding impacted the Mitigating Systems cornerstone and was of very low safety significance or Green because the finding was not a design deficiency or loss of function. Specifically, the inspectors identified the problem prior to the licensee incurring actual risk exposure time. The inspectors reviewed IMC 0310 and determined the cause of the finding involved the cross-cutting area of human performance, the component of resources, and the aspect of adequate emergency equipment, H.2(d), because the licensee failed to ensure the new fire hoses would not impact safety-related components such as ECCS sump screens during post-LOCA conditions.
Enforcement:
10 CFR Part 50, Appendix B, Criterion V, states in part that activities affecting quality shall be accomplished by documented procedures. Contrary to the above, on November 28, 2012, the licensee failed to adequately accomplish procedure ES-0419 to evaluate new fire hoses remaining in the reactor building for post-LOCA conditions. Because this violation was of very low safety significance and was entered into the licensee's corrective action program as CR-12-05730, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000395/2013004-03, Failure to Adequately Evaluate New Fire Hoses for Post-Loss of Coolant Accident Conditions.
4OA6 Meetings, Including Exit
On October 29, 2013, the resident inspectors presented the integrated inspection report results to Mr. T. Gatlin and other members of the licensee staff. The licensee acknowledged the results of these inspections. The inspectors confirmed that inspection activities discussed in this report did not contain proprietary material.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- J. Archie, Senior Vice President, Nuclear Operations
- A. Barbee, Director, Nuclear Training
- M. Browne, Manager, Quality Systems
- M. Coleman, Manager, Health Physics and Safety Services
- G. Douglass, Manager, Nuclear Protection Services
- T. Gatlin, Vice President, Nuclear Operations
- J. Wasieczko, Manager, Organization Development and Performance
- M. Harmon, Manager, Chemistry Services
- R. Haselden, General Manager, Organizational / Development Effectiveness
- R. Justice, Manager, Nuclear Operations
- G. Lippard, General Manager, Nuclear Plant Operations
- M. Mosley, Manager, Nuclear Training
- D. Perez, Supervisor, Health Physics - Technical Support
- S. Reese, Specialist, Nuclear Licensing
- J. Rinehart, Supervisor, Health Physics - Field Operations
- M. Roberts, Supervisor, Health Physics II, New Plant, Environmental, Rad Waste
- D. Shue, Manager, Maintenance Services
- W. Stuart, General Manager, Engineering Services
- B. Thompson, Manager, Nuclear Licensing
- D. Weir, Manager, Plant Support Engineering
- B. Wetmore, Design Engineering
- R. Williamson, Manager, Emergency Planning
- S. Zarandi, General Manager, Nuclear Support Services
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000395/2013004-01 NCV Failure to Evaluate Radiological Conditions (Section 2RS1)
- 05000395/2013004-02 FIN Failure to Develop Adequate Corrective Action for a Condition Adverse to Quality Involving the Loss of the Alternate AC Transformer (Section 4OA2.3)
- 05000395/2013004-03 NCV Failure to Adequately Evaluate New Fire Hoses for Post-
Loss of Coolant Accident Conditions (Section 4OA5.2)
Closed
- 05000395/2012-002-01 LER Seismically Qualified RWST Aligned to Non-Seismic Piping (Section 4OA3.1)
- 05000395/2012005-02 URI Post-LOCA impact of Replacement Fire Hoses on the ECCS Sump (Section 4OA5.2)