IR 05000382/2017004
| ML18038B043 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 02/07/2018 |
| From: | Geoffrey Miller NRC/RGN-IV/DRP/RPB-D |
| To: | Dinelli J Entergy Operations |
| Geoffrey Miller | |
| References | |
| IR 2017004 | |
| Download: ML18038B043 (52) | |
Text
bruary 7, 2018
SUBJECT:
WATERFORD STEAM ELECTRIC STATION, UNIT 3 - NRC INTEGRATED INSPECTION REPORT 05000382/2017004
Dear Mr. Dinelli:
On December 31, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Waterford Steam Electric Station, Unit 3. On January 11, 2018, the NRC inspectors discussed the results of this inspection with you and other members of your staff.
The results of this inspection are documented in the enclosed report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
Both 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.a of the Enforcement Policy.
If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; the Director, Office of Enforcement; and the NRC resident inspector at the Waterford Steam Electric Station, Unit 3.
If you disagree with a cross-cutting aspect assignment 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC resident inspector at the Waterford Steam Electric Station, Unit 3. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Geoffrey Miller, Branch Chief Project Branch D Division of Reactor Projects Docket No. 50-382 License No. NPF-38 Enclosure:
Inspection Report 05000382/2017004 w/ Attachments:
1. Supplemental Information 2. Public Radiation Safety Inspection Request for Information
SUNSI Review: ADAMS: Non-Publicly Available Non-Sensitive Keyword:
By: JDixon/dll Yes No Publicly Available Sensitive NRC-002 OFFICE SRI:DRP/D RI:DRP/D RIII/DRS BC:DRS/EB1 BC:DRS/EB2 BC:DRS/PSB2 NAME FRamirez CSpeer JCorujo-Sandin TFarnholtz GWerner HGepford SIGNATURE /RA/ /RA/ /RA/E /RA/ /RA/ /RA/
DATE 02/06/2018 02/04/2018 02/06/2018 01/31/2018 01/31/2018 02/01/18 OFFICE BC:DRS/OB TL:DRS/IPAT SPE:DRP/D BC:DRP/D NAME VGaddy THipschman JDixon GMiller SIGNATURE /RA/ /RA/ /RA/ /RA/
DATE 01/31/18 02/04/2018 02/06/2018 02/07/18
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000382 License: NPF-38 Report: 05000382/2017004 Licensee: Entergy Operations, Inc.
Facility: Waterford Steam Electric Station, Unit 3 Location: 17265 River Road Killona, LA 70057 Dates: October 1 through December 31, 2017 Inspectors: F. Ramírez, Senior Resident Inspector C. Speer, Resident Inspector P. Elkmann, Senior Emergency Preparedness Inspector G. Hansen, Senior Emergency Preparedness Inspector S. Hedger, Emergency Preparedness Inspector J. Corujo-Sandín, Reactor Inspector N. Okonkwo, Reactor Inspector L. Carson II, Senior Health Physicist N. Greene, PhD, Health Physicist S. Money, Health Physicist C. Alldredge, Health Physicist Approved Geoffrey Miller By: Chief, Project Branch D Division of Reactor Projects Enclosure
SUMMARY
IR 05000382/2017004; 10/01/2017 - 12/31/2017; Waterford Steam Electric Station, Unit 3;
Surveillance Testing, Problem Identification and Resolution.
The inspection activities described in this report were performed between October 1 and December 31, 2017, by the resident inspectors at Waterford Steam Electric Station, Unit 3, inspectors from the NRCs Region IV office, and other NRC offices. Two findings of very low safety significance (Green) are documented in this report. Both of these findings involved violations of NRC requirements. The significance of inspection findings is indicated by their color (i.e., Green, greater than Green, White, Yellow, or Red), determined using Inspection Manual Chapter 0609, Significance Determination Process, dated April 29, 2015. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, Aspects within the Cross-Cutting Areas, dated December 4, 2014. Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, dated July 2016.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XI, Test Control, for the licensees failure to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, the licensee failed to properly perform testing of the containment spray pump A in accordance with site Procedure OP-903-035, Containment Spray Pump Operability Check,
Revision 25. The licensee entered the issue into the corrective action program as Condition Reports CR-WF3-2017-09109 and CR-WF3-2017-09207. The licensees immediate corrective actions included performing an operability evaluation for the containment spray pump and determining the component remained operable.
The performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern.
Specifically, the licensees failure to perform pump testing in accordance with site procedures or the Operation and Maintenance Code could prevent identification of degrading performance adversely affecting the pumps capability to respond to an initiating event. The failure to correct the performance deficiency could also result in other safety-related pumps being incorrectly tested and failure of the licensee to identify the component might be degraded, deficient, and/or inoperable. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process, dated April 29, 2015. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated October 7, 2016, instructed the inspectors to use Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. Using Appendix A, Exhibit 2 - Mitigating Systems Screening Questions, the finding screened as Green because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, but the structure, system, or component maintained its operability.
The finding had a training cross-cutting aspect in the human performance area because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and did not instill nuclear safety values.
Specifically, licensee staff was under the incorrect impression that inservice testing could be repeated until acceptable results were obtained without the need to evaluate the reason for the failure [H.9]. (Section 1R22)
- Green.
The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33 for the licensees failure to properly perform maintenance on a safety-related component in accordance with site procedures. Specifically, the licensee failed to check the orientation of the operating springs for reactor trip circuit breaker 2 following maintenance as required per Procedure ME-004-155, Reactor Trip Switchgear, Revision 308. As a result, one of the two springs became loose, dropped into the breaker operating mechanism, and caused a breaker failure that was discovered during testing. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-05046. The licensees corrective actions included an extent of condition inspection of the other breakers in the reactor trip switchgear to ensure the springs were properly installed, and the addition of a critical step to Procedure ME-004-155, requiring an independent verification of the spring orientation following breaker maintenance.
The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the breaker was placed in service while incorrectly assembled which resulted in the breaker failing to open. The failure decreased the redundancy of the reactor trip circuit breakers and placed the unit in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown technical specification limiting condition for operation action statement. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609,
Significance Determination Process, dated April 29, 2015. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated October 7, 2016, instructed the inspectors to use Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. Using Appendix A, Exhibit 2 - Mitigating Systems Screening Questions, the finding screened as Green because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The finding had an avoid complacency cross-cutting aspect in the area of human performance because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, even with the different tools available, following each instance of maintenance on the reactor trip circuit breaker, the workers failed to recognize the improper installation of the breaker spring [H.12]. (Section 4OA2)
PLANT STATUS
The Waterford Steam Electric Station, Unit 3, began the inspection period at 100 percent power and maintained 100 percent power for the duration of the inspection period.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04 Equipment Alignment
.1 Partial Walk-Down
a. Inspection Scope
The inspectors performed partial system walk-downs of the following risk-significant systems:
- October 10, 2017, main switchgear ventilation train B with train A out of service for maintenance
- November 7, 2017, high pressure safety injection train B with train A out of service for planned maintenance
- December 16, 2017, essential chiller B following train realignment The inspectors reviewed the licensees procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the systems or trains were correctly aligned for the existing plant configuration.
These activities constituted three partial system walk-down samples, as defined in Inspection Procedure 71111.04.
b. Findings
No findings were identified.
.2 Complete Walk-Down
a. Inspection Scope
On November 28, 2017, the inspectors performed a complete system walk-down inspection of the emergency feedwater system. The inspectors reviewed the licensees procedures and system design information to determine the correct emergency feedwater system lineup for the existing plant configuration. The inspectors also reviewed outstanding work orders, open condition reports, temporary modifications, and other open items tracked by the licensees operations and engineering departments.
The inspectors then visually verified that the system was correctly aligned for the existing plant configuration.
These activities constituted one complete system walk-down sample, as defined in Inspection Procedure 71111.04.
b. Findings
No findings were identified.
1R05 Fire Protection
Quarterly Inspection
a. Inspection Scope
The inspectors evaluated the licensees fire protection program for operational status and material condition. The inspectors focused their inspection on four plant areas important to safety:
- October 4, 2017, west wing area, Fire Area RAB 25
- November 29, 2017, -35 reactor auxiliary building general area containing emergency feedwater pump AB, Fire Area RAB 39-001
- December 28, 2017, fire water pump house, Fire Area FWPH For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensees fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions.
These activities constituted four quarterly inspection samples, as defined in Inspection Procedure 71111.05.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
.1 Review of Licensed Operator Requalification
a. Inspection Scope
On October 19, 2017, the inspectors observed an evaluated simulator scenario performed by an operating crew. The inspectors assessed the performance of the operators and the evaluators critique of their performance.
These activities constituted completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
.2 Review of Licensed Operator Performance
a. Inspection Scope
On December 18, 2017, the inspectors observed the performance of on-shift licensed operators in the plants main control room. At the time of the observations, the plant was in a period of heightened activity due to testing of emergency diesel generator B and associated engineered safety feature actuations. The inspectors observed the operators performance of the following activities:
- Communications with field personnel
- Pump and valve manipulations
- Alarm response
- Crew briefs In addition, the inspectors assessed the operators adherence to plant procedures, including Procedure EN-OP-115, Conduct of Operations, Revision 23, and other operations department policies.
These activities constituted completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
.1 Routine Maintenance Effectiveness
a. Inspection Scope
On December 15, 2017, the inspectors reviewed one instance of degraded performance of safety-related structures, systems, and components (SSCs) on the sites spent fuel pool cooling and purification system.
The inspectors reviewed the extent of condition of possible common cause SSC failures and evaluated the adequacy of the licensees corrective actions. The inspectors reviewed the licensees work practices to evaluate whether these may have played a role in the degradation of the SSCs. The inspectors assessed the licensees characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.
On November 21, 2017, the inspectors reviewed the licensees periodic evaluation required by 10 CFR 50.65(a)(3) that evaluates performance and condition monitoring activities, and associated goals and preventive maintenance for SSCs. The inspectors verified that the periodic evaluation had been completed within the time constraints of the Maintenance Rule, and that the licensee had reviewed its 10 CFR 50.65(a)(1) goals, 10 CFR 50.65(a)(2) performance criteria, monitoring, and preventive maintenance activities, and effectiveness of corrective actions. In addition, the inspectors verified that industry operating experience had been taken into account where practical and the licensee made appropriate adjustments as a result of the periodic evaluation.
These activities constituted completion of two maintenance effectiveness samples, as defined in Inspection Procedure 71111.12.
b. Findings
No findings were identified.
.2 Quality Control
a. Inspection Scope
On December 22, 2017, the inspectors reviewed the licensees quality control activities through a review of parts installed in the emergency feedwater system that were purchased as commercial-grade parts but were dedicated prior to installation in a quality-grade application.
These activities constituted completion of one quality control sample, as defined in Inspection Procedure 71111.12.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
On November 3, 2017, the inspectors reviewed a risk assessment performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to planned Yellow risk due to transformer yard work and planned auxiliary component cooling water maintenance work window.
The inspectors verified that this risk assessment was performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensees risk assessment and verified that the licensee implemented appropriate risk management actions based on the results of the assessments.
The inspectors also observed portions of three emergent work activities that had the potential to cause an initiating event and to affect the functional capability of mitigating systems:
- October 19, 2017, emergent work related to the inoperability of main feedwater isolation valve 2
- November 16, 2017, emergent work and Yellow risk due to missed emergency diesel generator surveillance
- November 27, 2017, emergent work related to elevated vibrations on main feedwater pump B The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on unaffected SSCs.
These activities constituted completion of four maintenance risk assessments and emergent work control inspection samples, as defined in Inspection Procedure 71111.13.
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments
a. Inspection Scope
The inspectors reviewed three operability determinations that the licensee performed for degraded or nonconforming SSCs:
- October 31, 2017, operability determination of charging pump AB after pulsation dampener pressure was found low
- November 15, 2017, operability determination of high pressure safety injection cold legs 1A, 1B, and 2A degraded flow reading
- December 29, 2017, operability determination of essential chiller B following an unexpected trip The inspectors reviewed the timeliness and technical adequacy of the licensees evaluations. Where the licensee determined the degraded SSC to be operable, the inspectors verified that the licensees compensatory measures were appropriate to provide reasonable assurance of operability or functionality. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability of the degraded SSC.
These activities constituted completion of three operability review samples, as defined in Inspection Procedure 71111.15.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed four post-maintenance testing activities that affected risk-significant SSCs:
- October 13, 2017, low pressure safety injection system, train A, following planned maintenance
- October 31, 2017, control room ventilation air handling unit and emergency filtration unit, train A, following planned maintenance
- November 7, 2017, high pressure safety injection, train A, following planned maintenance
- November 30, 2017, emergency feedwater header B to steam generator 2 primary flow control valve following planned maintenance The inspectors reviewed licensing- and design-basis documents for the SSCs and the maintenance and post-maintenance test procedures. The inspectors observed the performance of the post-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected SSCs.
These activities constituted completion of four post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors observed five risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the SSCs were capable of performing their safety functions:
Reactor coolant system leak detection tests:
- November 17, 2017, reactor coolant system increased unidentified leakage Other surveillance tests:
- October 5, 2017, auxiliary component cooling water pump A
- October 17, 2017, emergency feedwater loop B flow calibration
- November 6, 2017, emergency diesel generator A monthly operability run
- November 8, 2017, containment spray pump A comprehensive test The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the tests satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected SSCs following testing.
These activities constituted completion of five surveillance testing inspection samples, as defined in Inspection Procedure 71111.22.
b. Findings
Introduction.
The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, the licensee failed to properly perform testing of containment spray pump A in accordance with site Procedure OP-903-035, Containment Spray Pump Operability Check, Revision 25.
Description.
On November 8, 2017, the licensee performed a comprehensive pump test of containment spray pump A, in accordance with Procedure OP-903-035, Containment Spray Pump Operability Check, Revision 25. This procedure is used, among other things, to satisfy the sites inservice testing program. Many of the procedural steps stated in Procedure OP-903-035 are obtained from code requirements established by the code of record. For this particular activity, the inservice testing was implemented in accordance with the American Society of Mechanical Engineers Operation and Maintenance (ASME OM) Code 2001, addenda through 2003, which was the code of record at the time of the inspection.
Comprehensive pump testing is a biennial test required by the licensees inservice testing program and the ASME OM Code. To perform the test, the licensee declares the system inoperable, aligns the system to permit full flow, and then adjusts backpressure via a manually operated valve until measured pump flow is within an allowable band.
Procedure OP-903-035 establishes a desired nominal pump flow value of 2,015 gpm and establishes an acceptable band of 2,000 - 2,025 gpm. Once flow is stabilized, the procedure requires that a number of parameters such as pump discharge pressure, suction pressure, and vibration data be measured and recorded. With this data, the licensee calculates the differential pressure as required by Procedure OP-903-035, records it, and compares it to the acceptance criteria. Procedure OP-903-035 establishes the differential pressure acceptance criteria as greater than or equal to 181.8 psid and less than or equal to 187.8 psid. Anything outside this band falls under the Required Action range. When test results fall under Required Action range, Procedure OP-903-035 provides instructions to declare the pump inoperable until cause of deviation has been determined and the condition corrected, to initiate a work order in accordance with Procedure EN-WM-100, Work Request (WR) Generation, Screening, and Classification, and to initiate a condition report in accordance with Procedure EN-LI-102, Corrective Action Program.
As stated in Procedure OP-903-035, the licensee recognizes that the tests differential pressure acceptance criteria is very restrictive, and flow should be established as close as possible to the nominal value. The inspectors noted that various factors influence the very restrictive acceptance criteria such as inservice testing program and design bases requirements. Compliance with these acceptance criteria ensures the pump will be able to perform its safety-related function during a postulated design bases event.
While observing the test, the inspectors noted the licensee set flow within the allowed band, measured discharge pressure and suction pressure, calculated differential pressure, and compared it to the applicable acceptance criteria. The licensee recognized the differential pressure fell into the Required Action range, which would also be considered a test failure. After consulting with control room personnel, the licensee adjusted system backpressure (i.e. adjusted pump flow), repeated the test, and again, the test failed. The licensee completed this iteration approximately four times, each time attempting to get closer to the nominal flow of 2015 gpm, until they obtained differential pressure test results that fell in the acceptable range. Once this was completed, the licensee then took vibration data. Repeating the test until satisfactory results were obtained was contrary to both the instructions provided by Procedure OP-903-035 and the ASME OM Code.
After completing the test, the inspectors questioned the licensee on the practice to repeat pump tests multiple times until obtaining passing results. The licensee explained it was not uncommon and not limited to the containment spray pump. Further, the inspectors learned that the licensee considered these test results satisfactory. In addition, the inspectors noted that even though it was required by Procedure OP-903-035, a condition report had not been generated documenting the problems in obtaining satisfactory differential pressure results, no determination regarding the cause of the deviation had been obtained, no work request had been initiated, and no operability assessment had been performed on the containment spray pump prior to restoring it to operable status. The inspectors also noticed that the official test record only documented the results of the last test (i.e successful results).
The inspectors concluded that the licensees failure to follow Procedure OP-903-035 could result in an invalid test; multiple flow adjustments to obtain acceptable differential pressure values (which could be considered unacceptable preconditioning based on the guidance provided in Part 9900: Technical Guidance, Maintenance - Preconditioning of Structures, Systems, and Components Before Determining Operability); and multiple retests which could mask pump degradation, a negative trend, or other SSC problems.
In addition, documenting only the last test (the passing results) prevents the corrective action program from evaluating the above-mentioned concerns. The objective of entering the corrective action program after obtaining results in the required action range is to ensure the licensee understands the reason the test is failing before proceeding (e.g. for engineering to evaluate the results). Retesting without evaluation effectively bypasses the corrective action program.
The licensee entered this issue into their corrective action program as CR-WF3-2017-09109 and CR-WF3-2017-09207. The licensee performed an operability evaluation, reviewed the results of the latest technical specification surveillance test, reviewed the trend results of the last few inservice tests and determined the containment spray pump A remained operable.
Analysis.
The inspectors determined that the failure to perform comprehensive and operability pump testing for the containment spray pump A, in accordance with site Procedure OP-903-035, Containment Spray Pump Operability Check, was a performance deficiency, which was reasonably within the licensees ability to foresee and correct. The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to perform inservice and operability pump testing in accordance with site procedures or the Operation and Maintenance Code could prevent identification of degrading performance adversely affecting the pumps capability to respond to an initiating event. The failure to correct the performance deficiency could also result in other safety-related pumps being incorrectly tested and failure of the licensee to identify the component might be degraded, deficient, and/or inoperable.
The inspectors screened the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated April 29, 2015.
IMC 0609, Attachment 4, Initial Characterization of Findings, dated October 7, 2016, instructed the inspectors to use Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. Using Appendix A, Exhibit 2 - Mitigating Systems Screening Questions, the finding screened as Green because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component, but the structure, system, or component maintained its operability.
The finding had a training cross-cutting aspect in the human performance area because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and did not instill nuclear safety values.
Specifically, licensee staff was under the incorrect impression that inservice testing could be repeated until acceptable results were obtained without the need to evaluate the reason for the failure [H.9].
Enforcement.
As required by 10 CFR Part 50, Appendix B, Criterion XI, Test Control, a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Contrary to the above, on November 8, 2017, for quality related components to which Appendix B applies, the licensee failed to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily inservice was identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
Specifically, while performing Procedure OP-903-035, Containment Spray Pump Operability Check, Revision 25, which is used to demonstrate that the safety-related containment spray pump is capable of meeting its safety-related function under design bases conditions, the licensee failed to properly collect and record test data and take required corrective actions when the test results fell in the Required Action range as specified in steps 6.4, and 7.1.19 through 7.1.21.1 of OP-903-035. As a result, the licensee failed to perform the procedurally-required actions established to ensure test failures are evaluated prior to returning the containment spray system to operable status.
The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-09109 and CR-WF3-2017-09207. As part of immediate corrective actions the licensee performed an operability evaluation for the containment spray pump and determined the component remained operable. Because this violation was of very low safety significance (Green) and was entered into the licensees corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000382/2017004-01, Failure to Follow Testing Requirements for the Containment Spray Pump A)
Cornerstone: Emergency Preparedness
1EP1 Exercise Evaluation
a. Inspection Scope
The inspectors observed the December 5, 2017, biennial emergency preparedness exercise to verify the exercise acceptably tested the major elements of the emergency plan and provided opportunities for the emergency response organization to demonstrate key skills and functions. The scenario demonstrated the licensees capability to implement its emergency plan by simulating:
- Elevated concentrations of radioactive materials in the reactor coolant because of leaking fuel assemblies (as an initial condition of the exercise)
- Physical damage to the exterior of the fuel handling building
- Unexpected insertion of reactor control rods into the core
- Leaks from the instrument air system
- Failures on two charging pumps
- A failure of the reactor protection system to trip the reactor with successful manual actions
- A primary to secondary leak on a steam generator leading to a tube rupture
- An unfiltered unmonitored radiological release to the environment from a steam leak on the main steam isolation valves During the exercise, the inspectors observed activities in the control room simulator, technical support center, operations support center, and the emergency operations facility. The inspectors focused their evaluation of the licensees performance on the risk-significant activities of event classification, offsite notification, recognition of offsite dose consequences, and development of protective action recommendations.
The inspectors also assessed recognition of, and response to, abnormal and emergency plant conditions, the transfer of decision-making authority and emergency function responsibilities between facilities, on-site and offsite communications, protection of emergency workers, emergency repair evaluation and capability, and the overall implementation of the emergency plan to protect public health and safety. The inspectors reviewed the current revision of the facility emergency plan, emergency plan implementing procedures associated with operation of the licensees emergency response facilities, procedures for the performance of associated emergency functions, and other documents as listed in the attachment to this report.
The inspectors attended the post-exercise critiques in each emergency response facility to evaluate the initial licensee self-assessment of exercise performance. The inspectors also attended a formal presentation of critique items to plant management conducted Tuesday, December 12, 2017.
The inspectors reviewed the scenarios of previous licensee exercises and drills conducted between January 2016, and November 2017, to determine whether the December 5, 2017, exercise was independent and avoided participant preconditioning, in accordance with the requirements of 10 CFR Part 50, Appendix E, IV.F(2)(g). The inspectors also compared observed exercise performance with corrective action program entries and After-Action reports for drills and exercises conducted between January 2016 and November 2017 to determine whether previously-identified weaknesses had been corrected in accordance with the requirements of 10 CFR 50.47(b)(14), and 10 CFR Part 50, Appendix E, IV.F.
The inspectors also discussed exercise performance with staff at Federal Emergency Management Agency (FEMA) Region VI to determine whether the exercise adequately supported the FEMA exercise evaluation objectives.
These activities constituted one exercise evaluation sample, as defined in Inspection Procedure 71114.01.
b. Findings
No findings were identified.
1EP6 Drill Evaluation
Training Evolution Observation
a. Inspection Scope
On October 19, 2017, the inspectors observed simulator-based licensed operator training that included implementation of the licensees emergency plan. The inspectors verified that the licensees emergency classifications, off-site notifications, and protective action recommendations were appropriate and timely. The inspectors verified that any emergency preparedness weaknesses were appropriately identified by the evaluators and entered into the corrective action program for resolution.
These activities constituted completion of one training observation sample, as defined in Inspection Procedure 71114.06.
b. Findings
No findings were identified.
1EP8 Exercise Evaluation - Scenario Review
a. Inspection Scope
The licensee submitted the preliminary exercise scenario for the December 5, 2017, biennial exercise to the NRC on October 5, 2017, in accordance with the requirements of 10 CFR Part 50, Appendix E, IV.F(2)(b). The inspectors performed an in-office review of the proposed scenario to determine whether it would acceptably test the major elements of the licensees emergency plan, and provide opportunities for the emergency response organization to demonstrate key skills and functions. The inspectors also discussed the preliminary scenario with staff at FEMA Region VI to determine whether the proposed exercise would support the FEMA exercise evaluation objectives.
These activities constituted completion of one exercise evaluation sample, as defined in Inspection Procedure 71114.08.
b. Findings
No findings were identified.
RADIATION SAFETY
Cornerstones: Public Radiation Safety and Occupational Radiation Safety
2RS5 Radiation Monitoring Instrumentation
a. Inspection Scope
The inspectors evaluated the accuracy and operability of the radiation monitoring equipment used by the licensee to monitor areas, materials, and workers to ensure a radiologically safe work environment. This evaluation included equipment used to monitor radiological conditions related to normal plant operations, anticipated operational occurrences, and conditions resulting from postulated accidents. The inspectors interviewed licensee personnel, walked down various portions of the plant, and reviewed licensee performance associated with radiation monitoring instrumentation, as described below:
- The inspectors performed walk downs and observations of selected plant radiation monitoring equipment and instrumentation, including portable survey instruments, area radiation monitors, continuous air monitors, personnel contamination monitors, portal monitors, and small article monitors. The inspectors assessed material condition and operability, evaluated positioning of instruments relative to the radiation sources or areas they were intended to monitor, and verified performance of source checks and calibrations.
- The inspectors evaluated the calibration and testing program, including laboratory instrumentation, whole body counters, post-accident monitoring instrumentation, portal monitors, personnel contamination monitors, small article monitors, portable survey instruments, area radiation monitors, electronic dosimetry, air samplers, and continuous air monitors.
- The inspectors assessed problem identification and resolution for radiation monitoring instrumentation. The inspectors reviewed audits, self-assessments, and corrective action program documents to verify problems were being identified and properly addressed for resolution.
These activities constituted completion of the three required samples of radiation monitoring instrumentation, as defined in Inspection Procedure 71124.05.
b. Findings
No findings were identified.
2RS6 Radioactive Gaseous and Liquid Effluent Treatment
a. Inspection Scope
The inspectors evaluated whether the licensee maintained gaseous and liquid effluent processing systems and properly mitigated, monitored, and evaluated radiological discharges with respect to public exposure. The inspectors verified that abnormal radioactive gaseous or liquid discharges and conditions, when effluent radiation monitors are out-of-service, were controlled in accordance with the applicable regulatory requirements and licensee procedures. The inspectors verified that the licensees quality control program ensured radioactive effluent sampling and analysis adequately quantified and evaluated discharges of radioactive materials. The inspectors verified the adequacy of public dose projections resulting from radioactive effluent discharges. The inspectors interviewed licensee personnel and reviewed licensee performance in the following areas:
- During walk downs and observations of selected portions of the radioactive gaseous and liquid effluent equipment, the inspectors evaluated routine processing and discharge of effluents, including sample collection and analysis.
The inspectors observed equipment configuration and flow paths of selected gaseous and liquid discharge system components, effluent monitoring systems, filtered ventilation system material condition, and significant changes to effluent release points.
- Calibration and testing program for process and effluent monitors, including National Institute of Standards and Technology (NIST) traceability of sources, primary and secondary calibration data, channel calibrations, set-point determination bases, and surveillance test results.
- Sampling and analysis controls used to ensure representative sampling and appropriate compensatory sampling. Reviews included results of the inter-laboratory comparison program.
- Instrumentation and equipment, including effluent flow measuring instruments, air cleaning systems, and post-accident effluent monitoring instruments.
- Dose calculations for effluent releases. The inspectors reviewed a selection of radioactive liquid and gaseous waste discharge permits and abnormal gaseous or liquid tank discharges, and verified the projected doses were accurate. The inspectors also reviewed 10 CFR Part 61 analyses and methods used to determine which isotopes were included in the source term. The inspectors reviewed land use census results, offsite dose calculation manual changes, and significant changes in reported dose values from previous years.
- Problem identification and resolution for radioactive gaseous and liquid effluent treatment. The inspectors reviewed audits, self-assessments, and corrective action program documents to verify problems were being identified and properly addressed for resolution.
These activities constituted completion of the six required samples of the radioactive gaseous and liquid effluent treatment program, as defined in Inspection Procedure 71124.06.
b. Findings
No findings were identified.
2RS7 Radiological Environmental Monitoring Program
a. Inspection Scope
The inspectors evaluated whether the licensees radiological environmental monitoring program quantified the impact of radioactive effluent releases to the environment and sufficiently validated the integrity of the radioactive gaseous and liquid effluent release program. The inspectors also verified that the licensee continued to implement the voluntary Nuclear Energy Institute (NEI) Industrys Ground Water Protection Initiative (GPI). The inspectors reviewed or observed the following items:
- The inspectors observed selected air sampling and dosimeter monitoring stations, sampler station modifications, and the collection and preparation of environmental samples. The inspectors reviewed calibration and maintenance records for selected air samplers, composite water samplers, and environmental sample radiation measurement instrumentation, and inter-laboratory comparison program results. The inspectors reviewed selected events documented in the annual environmental monitoring report and significant changes made by the licensee to the offsite dose calculation manual as the result of changes to the land census. The inspectors evaluated the operability, calibration, and maintenance of meteorological instruments and assessed the meteorological dispersion and deposition factors. The inspectors verified the licensee had implemented sampling and monitoring programs sufficient to detect leakage from structures, systems, or components with credible mechanism for licensed material to reach ground water and reviewed changes to the licensees written program for identifying and controlling contaminated spills/leaks to groundwater.
- Groundwater protection initiative implementation, including assessment of groundwater monitoring results, identified leakage or spill events and entries made into 10 CFR 50.75(g) records, licensee evaluations of the extent of the contamination and the radiological source term, and reports of events associated with spills, leaks, and groundwater monitoring results.
- Problem identification and resolution for the radiological environmental monitoring program. The inspectors reviewed audits, self-assessments, and corrective action program documents to verify problems were being identified and properly addressed for resolution.
These activities constituted completion of the three required samples of the radiological environmental monitoring program, as defined in Inspection Procedure 71124.07.
b. Findings
No findings were identified.
2RS8 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage,
and Transportation (71124.08)
a. Inspection Scope
The inspectors evaluated the effectiveness of the licensees programs for processing, handling, storage, and transportation of radioactive material. The inspectors interviewed licensee personnel and reviewed the following items:
- Radioactive material storage, including waste storage areas including container labeling/marking and monitoring containers for deformation or signs of waste decomposition.
- Radioactive waste system, including walk-downs of the accessible portions of the radioactive waste processing systems and handling equipment. The inspectors also reviewed or observed changes made to the radioactive waste processing systems, methods for dewatering and waste stabilization, waste stream mixing methodology, and waste processing equipment that was not operational or abandoned in place.
- Waste characterization and classification, including radio-chemical sample analysis results for radioactive waste streams and use of scaling factors and calculations to account for difficult-to-measure radionuclides, and processes for waste classification including use of scaling factors and 10 CFR Part 61 analyses.
- Shipment preparation, including packaging, surveying, labeling, marking, placarding, vehicle checking, driver instructing, and preparation of the disposal manifests.
- Shipping records for Low Specific Activity (LSA-I, LSA-II, and LSA-III), Surface Contaminated Objects (SCO-I and SCO-II), Type A, or Type B, radioactive material or radioactive waste shipments.
- Problem identification and resolution for radioactive solid waste processing and radioactive material handling, storage, and transportation. The inspectors reviewed audits, self-assessments, and corrective action program documents to verify problems were being identified and properly addressed for resolution.
These activities constituted completion of the six required samples of the radioactive solid waste processing, and radioactive material handling, storage, and transportation program, as defined in Inspection Procedure 71124.08.
b. Findings
No findings were identified.
OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security
4OA1 Performance Indicator Verification
.1 Reactor Coolant System Specific Activity (BI01)
a. Inspection Scope
The inspectors reviewed the licensees reactor coolant system chemistry sample analyses for the period of October 1, 2016, through September 30, 2017, to verify the accuracy and completeness of the reported data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data.
These activities constituted verification of the reactor coolant system specific activity performance indicator, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
.2 Reactor Coolant System Identified Leakage (BI02)
a. Inspection Scope
The inspectors reviewed the licensees records of reactor coolant system identified leakage for the period of October 1, 2016, through September 30, 2017, to verify the accuracy and completeness of the reported data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data.
These activities constituted verification of the reactor coolant system leakage performance indicator, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
.3 Drill/Exercise Performance (EP01)
a. Inspection Scope
The inspectors reviewed the licensees evaluated exercises, emergency plan implementations, and selected drill and training evolutions that occurred between April 2016, and September 2017, to verify the accuracy of the licensees data for classification, notification, and protective action recommendation (PAR) opportunities.
The inspectors reviewed a sample of the licensees completed classifications, notifications, and PARs to verify their timeliness and accuracy. The inspectors used Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data. The specific documents reviewed are described in the attachment to this report.
These activities constituted verification of the drill/exercise performance indicator, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
.4 Emergency Response Organization Drill Participation (EP02)
a. Inspection Scope
The inspectors reviewed the licensees records for participation in drill and training evolutions between April 2016, and September 2017, to verify the accuracy of the licensees data for drill participation opportunities. The inspectors verified that all members of the licensees emergency response organization (ERO) in the identified key positions had been counted in the reported performance indicator data. The inspectors reviewed the licensees basis for reporting the percentage of ERO members who participated in a drill. The inspectors reviewed drill attendance records and verified a sample of those reported as participating. The inspectors used Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data. The specific documents reviewed are described in the attachment to this report.
These activities constituted verification of the emergency response organization drill participation performance indicator, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
.5 Alert and Notification System Reliability (EP03)
a. Inspection Scope
The inspectors reviewed the licensees records of alert and notification system tests conducted between April 2016, and September 2017, to verify the accuracy of the licensees data for siren system testing opportunities. The inspectors reviewed procedural guidance on assessing alert and notification system opportunities and the results of periodic alert and notification system operability tests. The inspectors used Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to determine the accuracy of the reported data. The specific documents reviewed are described in the attachment to this report.
These activities constituted verification of the alert and notification system reliability performance indicator, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution
.1 Routine Review
a. Inspection Scope
Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensees corrective action program and periodically attended the licensees condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified. The inspectors also reviewed the licensees problem identification and resolution activities during the performance of the other inspection activities documented in this report.
b. Findings
No findings were identified.
.2 Semiannual Trend Review
a. Inspection Scope
The inspectors reviewed the licensees corrective action program, performance indicators, system health reports, maintenance rule database, and other documentation to identify trends that might indicate the existence of a more significant safety issue. The inspectors verified that the licensee was taking corrective actions to address identified adverse trends.
These activities constituted completion of one semiannual trend review sample, as defined in Inspection Procedure 71152.
b. Observations and Assessments The inspectors reviewed condition reports associated with issues with the actuators of the control and isolation valves of the emergency feedwater system. The inspectors noted that the licensee is already tracking and resolving these issues in Condition Reports CR-WF3-2016-06706, and CR-WF3-2017-03195.
c. Findings
No findings were identified.
.3 Annual Follow-up of Selected Issues
a. Inspection Scope
The inspectors selected one issue for an in-depth follow-up:
- On November 13, 2017, the inspectors completed a review of a licensee adverse condition evaluation documented in Condition Report CR-WF3-2013-05046. The licensees evaluation was documented to assess the reactor trip circuit breaker 2 failure on June 11, 2017, during a surveillance test.
The inspectors assessed the licensees problem identification threshold, cause analyses, and extent of condition review. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions were adequate to correct the condition.
These activities constituted completion of one annual follow-up sample, as defined in Inspection Procedure 71152.
b. Findings
Introduction.
The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33 for the licensees failure to properly perform maintenance on a safety-related component in accordance with site procedures. Specifically, the licensee failed to check the orientation of the operating springs for reactor trip circuit breaker 2 following maintenance as required per Procedure ME-004-155, Reactor Trip Switchgear, Revision 306. As a result, one of the two springs became loose, dropped into the breaker operating mechanism and caused a breaker failure that was discovered during testing.
Description.
On June 11, 2017, the licensee was testing reactor trip circuit breakers 2 and 6 in accordance with Procedure OP-903-107, Plant Protection System Channel A, B, C, D Functional Test, Revision 310. The reactor trip switchgear consists of eight electrically operated air circuit breakers for interrupting power to the control element drive mechanism coils and one electrically operated bus tie breaker. The reactor trip circuit breakers actuate upon a trip demand initiated by the reactor protection system.
During the test, reactor trip circuit breaker 6 opened as expected; however, reactor trip circuit breaker 2 remained closed. Following the reactor trip circuit breaker 2 failure, at 8:47 a.m., the licensee entered Technical Specification 3.3.1, Action 5, which is a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limiting condition for operation (LCO) to isolate reactor trip circuit breaker 2 or be in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Operations personnel opened, racked out, and removed reactor trip circuit breaker 2, complying with the LCO action at 10:47 a.m. Since the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowed outage time had elapsed, the licensee commenced preparations for performing a plant shutdown to hot standby, as required by Technical Specification 3.3.1. However, once in compliance with Technical Specification 3.3.1, operations personnel halted the preparations for a plant shutdown.
Maintenance personnel performed troubleshooting for reactor trip circuit breaker 2 and found that one of the two breaker operating springs had become loose, dropped into the operating mechanism, and prevented the breaker from opening. After verifying the spare breakers closing springs were installed correctly and testing it, maintenance personnel replaced the failed circuit breaker with the spare breaker. In addition, maintenance personnel performed an extent of condition inspection of the other breakers in the reactor trip switchgear to ensure that their springs were installed correctly.
The licensees adverse condition analysis concluded that the cause for the spring becoming loose and disengaging from the breaker was improper installation. Reactor trip circuit breaker 2 was overhauled on October 1, 2003, under Work Order 24317 and Procedure ME-004-154, Reactor Trip Switchgear Overhaul Procedure, Revision 0.
During this maintenance activity, the spring was installed incorrectly. Specifically, the left spring was incorrectly installed with respect to orientation, which caused it to completely disengage and fall. Additionally, the preventive maintenance frequency for reactor trip circuit breakers is 3 years and is performed in accordance with Procedure ME-004-155. Both Procedure ME-004-154 and Procedure ME-004-155 contain steps that require checking for spring orientation following any type of maintenance. In addition to the required steps to check for orientation, the procedure includes a reference diagram to compare the springs in the field to detect improper orientation. The last time preventive maintenance had been performed on reactor trip circuit breaker 2 was April 14, 2015. Even though there were multiple maintenance activities between 2003 and 2015, maintenance personnel failed to discover the improper spring orientation during these work activities. The last surveillance testing for reactor trip circuit breaker 2, per Procedure OP-903-006, Reactor Trip Circuit Breaker Test, Revision 11, was satisfactorily completed on May 28, 2017, where the breaker cycled as required.
Analysis.
The inspectors concluded that the licensees failure to properly check the orientation of the operating springs for reactor trip circuit breaker 2 following maintenance as required per site Procedure ME-004-155 was a performance deficiency which was reasonably within the licensees ability to foresee and correct. The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and its objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the breaker was placed in service while it was assembled incorrectly. As a result, the failure decreased the redundancy of the reactor trip circuit breakers and placed the unit in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown technical specification LCO action statement.
The inspectors screened the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated April 29, 2015.
IMC 0609, Attachment 4, Initial Characterization of Findings, dated October 7, 2016, instructed the inspectors to use Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012. Using Appendix A, Exhibit 2 - Mitigating Systems Screening Questions, the finding screened as Green because the finding:
- (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component;
- (2) did not represent a loss of system and/or function;
- (3) did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and
- (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The finding had an avoid complacency cross-cutting aspect in the area of human performance because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
Specifically, even with the different tools available, following each instance of maintenance on the reactor trip circuit breaker, the workers failed to recognize the improper installation of the breaker spring [H.12].
Enforcement.
Technical Specification 6.8.1.a, requires, in part, that procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2. Appendix A to Regulatory Guide 1.33, Section 9.a, requires, in part, that, maintenance that can affect the performance of safety-related equipment be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. The licensee established, in part, Procedure ME-004-155, Reactor Trip Switchgear, Revision 306, to meet the Regulatory Guide 1.33 requirement. Step 9.3.9 of Procedure ME-004-155 states, inspect the springs for proper orientation.
Contrary to the above requirements, between October 2003 and June 2017, the licensee did not inspect the springs for proper orientation. Specifically, following breaker preventive maintenance, the licensee did not adequately complete the procedure Step 9.3.9 to verify the breaker springs for proper orientation. As a result, the left spring in reactor trip circuit breaker 2 became loose, dropped into the operating mechanism, and prevented the breaker from opening during a surveillance test on June 11, 2017. This failure placed the unit in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown technical specification LCO action statement. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-05046. The licensees corrective actions included an extent of condition inspection of the other breakers in the reactor trip switchgear to ensure the springs were properly installed, and the addition of a critical step to Procedure ME-004-155, requiring an independent verification of the spring orientation following breaker maintenance. Because this violation was of very low safety significance (Green) and was entered into the licensees corrective action program, this violation is being treated as a non-cited violation (NCV) in accordance with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000382/2017004-02, Failure to Perform Adequate Maintenance on a Reactor Trip Circuit Breaker)
4OA6 Meetings, Including Exit
Exit Meeting Summary
On November 2, 2017, the emergency preparedness inspectors discussed the in-office review of the preliminary scenario for the December 5, 2017, biennial exercise, submitted October 5, 2017, with Mr. J. Signorelli, Manager, Emergency Preparedness, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
On December 1, 2017, the radiation protection inspectors presented the radiation safety inspection results to Mr. J. Dinelli, Site Vice President, and other members of the licensee staff.
The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
On December 13, 2017, the emergency preparedness inspectors presented the results of the on-site inspection of the biennial emergency preparedness exercise conducted December 5, 2017, to Mr. D. Brenton, General Manager, Plant Operations, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
On January 11, 2018, the resident inspectors presented the inspection results to Mr. J. Dinelli, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- J. Bradley, Manager, Operations
- D. Breaud, Radiological Effluent Specialist, Chemistry
- D. Brenton, General Manager, Plant Operations
- L. Brown, Licensing Specialist, Regulatory Assurance
- T. Burnett, Director, Entergy Corporate Emergency Preparedness
- R. DeVoe, Licensing Specialist
- J. Dinelli, Site Vice President
- J. Frederickson, Support Supervisor, Radiation Protection
- D. James, Shipping Technician, Radiation Protection
- N. Justice, Specialist, Radiation Protection and Chemistry
- B. Lanka, Director, Engineering
- J. Lewis, Senior Project Manager, Entergy Corporate
- J. McBrayer, Acting Manager, Performance Improvement
- D. McLaren, Manager, Radiation Protection
- S. Meiklejohn, Senior Licensing Specialist
- P. Moritzky, Radwaste Operations Supervisor, Radiation Protection
- E. Neal, Superintendent, Radiation Protection
- J. Signorelli, Manager, Emergency Preparedness
Other Contacts
- D. Bordelon, Branch Chief, Technological Hazards, FEMA Region VI
- N. Williams, Chairperson, Radiological Assistance Committee, FEMA Region VI
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
Failure to Follow Testing Requirements for the Containment
- 05000382/2017004-01 NCV Spray Pump A (Section 1R22)
Failure to Perform Adequate Maintenance on a Reactor Trip
- 05000382/2017004-02 NCV Circuit Breaker (Section 4OA2)