IR 05000397/2013005
| ML14030A606 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/30/2014 |
| From: | Ryan Lantz NRC/RGN-IV/DRP/RPB-D |
| To: | Reddemann M Energy Northwest |
| Lantz R | |
| References | |
| IR-13-005 | |
| Download: ML14030A606 (58) | |
Text
January 30, 2014
SUBJECT:
COLUMBIA GENERATING STATION - NRC INTEGRATED INSPECTION REPORT 05000397/2013005
Dear Mr. Reddemann:
On December 31, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Columbia Generating Station. On January 6, 2014, the NRC inspectors discussed the results of this inspection with you and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
All of these findings involved violations of NRC requirements.
Further, inspectors documented two licensee-identified violations which were determined to be of very low safety significance in this report. 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, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Columbia Generating Station.
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 Regional Administrator, Region IV; and the NRC resident inspector at the Columbia Generating Station.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRC's U N IT E D S TA TE S NUC LEAR RE GULATOR Y C OM MI S SI ON R E G IO N I V 1600 EAST LAMAR BLVD AR L INGTON, TEXAS 76011-4511 Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Ryan E. Lantz, Branch Chief Project Branch D Division of Reactor Projects
Docket No: 50-397 License No: NPF-21
Enclosure: Inspection Report 05000397/2013005 w/ Attachment: Supplemental Information
cc w/ encl: Electronic Distribution
SUMMARY
IR 05000397/2013005; 09/22/2013 - 12/31/2013; Columbia Generating Station; Flood
Protection Measures; Surveillance Testing
The inspection activities described in this report were performed between September 22, 2013, and December 31, 2013, by the resident inspectors at Columbia Generating Station and seven inspectors from the NRCs Region IV office. Two findings of very low safety significance (Green) are documented in this report. All of these findings involved violations of NRC requirements. The significance of inspection findings is indicated by their color (Green, White,
Yellow, or Red), which is determined using Inspection Manual Chapter 0609, Significance Determination Process. Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, Components Within the Cross-Cutting Areas. Violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a non-cited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, for the licensees failure to translate the design of water resistant doors used to protect emergency core cooling rooms from internal flooding into procedures used to control those doors. The licensee entered this finding into their corrective action program as Action Request AR 298068298068
The performance deficiency was more than minor because it affected the procedure quality attribute of the mitigating systems cornerstone objective and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
This finding is of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical 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 />. This finding had a cross-cutting aspect in the area of human performance associated with the decision making component because the licensee failed to verify the validity of the underlying assumptions used in the stations flooding analysis and failed to identify possible unintended consequences when making changes to the barrier impairment procedure H.1(b)
(Section 1R06).
Cornerstone: Barrier Integrity
- Green.
The inspectors identified a non-cited violation of 10 CFR Part 50 Appendix B,
Criterion III, Design Control, for the licensees failure to translate the results of calculation NE-02-02-01, Control Room Boundary Leakage Limitation into allowed breach specifications for the control room ventilation boundary. This finding was entered into the licensees corrective action program as Action Request AR 298914298914
This performance deficiency was more than minor because it affected the design control attribute of the Barrier Integrity Cornerstone objective of providing reasonable assurance physical design barriers to protect the public from radionuclide releases caused by accidents or events. This finding is of very low safety significance (Green)because the finding only represents a degradation of the radiological barrier function provided for by the control room. The finding did not have a cross-cutting aspect because the performance deficiency occurred in early 2010 using a different process than currently exists and was therefore not reflective of current performance (Section 1R22).
Licensee-Identified Violations
Violations of very low safety significance that were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. These violations and associated corrective action tracking numbers are listed in Section 4OA7 of this report.
PLANT STATUS
The plant began the inspection period at 100 percent power. The plant operated at 100 percent power, with the exception of scheduled reductions in power to support maintenance and testing, for the remainder of the inspection period.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
.1 Readiness for Impending Adverse Weather Conditions
a. Inspection Scope
On September 30, 2013, the inspectors completed an inspection of the stations readiness for impending adverse weather conditions. The inspectors reviewed plant design features, the licensees procedures to respond to high winds, and the licensees implementation of these procedures. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant.
These activities constituted one sample of readiness for impending adverse weather conditions, as defined in Inspection Procedure 71111.01.
b. Findings
No findings were identified.
1R04 Equipment Alignment
.1 Partial Walkdown
a. Inspection Scope
The inspectors performed partial system walk-downs of the following risk-significant systems:
October 10, 2013, diesel fuel oil transfer system
October 10, 2013, ultimate heat sink
November 19, 2013, train A standby gas treatment system
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 were correctly aligned for the existing plant configuration.
These activities constituted three partial system walkdown samples as defined in Inspection Procedure 71111.04.
b. Findings
No findings were identified.
.2 Complete Walkdown
a. Inspection Scope
On December 18-20, 2013, the inspectors performed a complete system walk-down inspection of the division 1 and 2 125V DC distribution system. The inspectors reviewed the licensees procedures and system design information to determine the correct system lineup for the existing plant configuration. The inspectors also reviewed outstanding work orders, open condition reports, in-process design changes, 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
.1 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, 2013, Fire Area RC-2, cable spreading room
October 7, 2013, Fire Area R-21, reactor building 522 pipe space
October 16, 2013, Fire Areas RC-4, 5, 6, 7, 8, and 9 and RC-14, radioactive waste building vital island
November 19, 2013, Fire Area RC-13, emergency chiller area
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.
1R06 Flood Protection Measures
a. Inspection Scope
On November 19, 2013, the inspectors completed an inspection of the stations ability to mitigate flooding due to internal causes. After reviewing the licensees flooding analysis, the inspectors chose one plant area containing risk-significant structures, systems, and components that were susceptible to flooding:
Low pressure core spray pump room
The inspectors reviewed plant design features and licensee procedures for coping with internal flooding. The inspectors walked down the selected areas to inspect the design features, including the material condition of seals, drains, and flood barriers. The inspectors evaluated whether operator actions credited for flood mitigation could be successfully accomplished.
These activities constitute completion of one flood protection measures sample, as defined in Inspection Procedure 71111.06.
b. Findings
Introduction.
The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate the design of water resistant doors used to protect emergency core cooling rooms from internal flooding into procedures used to control those doors.
Description.
The inspectors reviewed the licensees strategies used to mitigate the effects of an internal flooding event, including the design of emergency core cooling system pump room water-resistant doors. Columbia Generating Station Final Safety Analysis Report, Section 3.4.1.5.2, Internal Flood Protection Measures, states that these doors are provided with seals that will minimize flooding between rooms even with significant hydrostatic pressure generated from flooding water levels up to 466 feet. In calculation ME-02-93-57, Effects of Stairwell Flooding on Adjacent Pump Rooms (422 feet 3 inches in Elevation), Revision 1, the licensee determined that the design of Columbia Generating Station is acceptable with respect to internal flooding because water-resistant doors will minimize flooding between adjacent pump rooms. A single water-resistant door is provided between emergency core cooling pump rooms and adjoining corridors, but two doors are provided when used between adjoining pump rooms. Double doors are necessary because the water-resistant doors are supplied with compression gaskets. For a single door arrangement, the flooding source would apply hydrostatic pressure to press the flood door away from its gasket which would allow the source of the internal flooding to communicate with adjoining pump rooms. In a double-door arrangement, the flooding source would always apply hydrostatic pressure to press at least one door into its gasket thereby sealing the room and confining the flooding source to a single emergency core cooling system pump room.
The inspectors compared the requirements of Procedure PPM 1.3.57, Barrier Impairment, Revision 29, to the design assumptions specified in calculation ME-02-93-57 and the FSAR. In that procedure, step 4.11.2.2 provides procedural requirements for Reactor Building 422 water-resistant doors and specified, in part, that door removal or other maintenance where the door(s) cannot be closed is a Mode 5 task unless adequate compensatory measures are implemented and justified in a 50.59 evaluation. This step does not apply to flood barriers with double watertight doors and where only one of the watertight doors is open.
The inspectors concluded that procedure PPM 1.3.57 would allow continuous operation with one of the double water-resistant flood doors between emergency core cooling system pump rooms out-of-service. In this configuration, the remaining water-resistant flood door would not be able to minimize flooding between pump rooms with hydrostatic pressure generated from water levels up to 466 feet. Operation in this configuration is therefore outside of the design specified in calculation ME-02-93-57 and Section 3.4.1.5.2 of the Columbia Generating Station Final Safety Analysis Report.
The licensee added the subject requirements in Step 4.11.2.c of PPM 1.3.57, Revision 26, effective October 7, 2010.
After becoming aware of this issue, the licensee initiated Action Request AR 298068298068to address the inadequacies associated with procedure PPM 1.3.57. The licensee also initiated Standing Order 12 on December 17, 2013, to provide additional operational guidance for water-resistant doors located between emergency core cooling system pump rooms.
The inspectors determined that the apparent cause of the performance deficiency was the failure to verify the assumption that one door between pump rooms would adequately resist hydraulic pressure to minimize flooding.
Analysis.
The failure to translate requirements specified in FSAR Section 3.4.1.5.2 and Calculation ME-02-93-57 into Procedure PPM 1.3.57 was a performance deficiency.
The performance deficiency was more than minor because it affected the procedure quality attribute of the mitigating systems cornerstone objective and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this performance deficiency resulted in guidance to operators which would allow continuous operation outside of the plants design basis. The inspectors performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power." Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined this finding is of very low safety significance (Green) because:
- (1) the finding was not a deficiency affecting the design or qualification of a mitigating system;
- (2) the finding did not represent a loss of system and/or function;
- (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and
- (4) the finding does not represent an actual loss of function of one or more non-technical 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 />. Because the apparent cause of this finding was the failure to verify the assumption that one door would adequately resist hydraulic pressure to minimize flooding, this finding has a cross-cutting aspect in the area of human performance associated with the decision making component, because the licensee did not use conservative assumptions in decision-making, and made a safety-significant decision without verifying the validity of underlying assumptions and without identifying possible unintended consequences H.1(b).
Enforcement.
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, on October 7, 2010, the licensee failed to translate applicable regulatory requirements and the Columbia Generating Station design basis into station procedures. Specifically, the licensee failed to translate the design of emergency core cooling system water-resistant doors as described in the Columbia Generating Station Final Safety Analysis Report, Section 3.4.1.5.2, Internal Flood Protection Measures, and Calculation ME-02-93-57, Effects of Stairwell Flooding on Adjacent Pump Rooms (422 ft 3 in Elevation), Revision 1, into procedure PPM 1.3.57, Barrier Impairment, Revision 26, Step 4.11.2.c. The licensee restored compliance by issuing Standing Order 12 on December 17, 2013, which prevented operation of these doors outside of the design basis. Because the finding is of very low safety significance (Green) and has been entered into the licensees corrective action program as Action Request AR 298068298068 this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a. of the NRC Enforcement Policy: NCV 05000397/2013005-01, Failure to Translate Internal Flooding Design into Station Procedures.
1R07 Heat Sink Performance
a. Inspection Scope
On December 19, 2013, the inspectors completed an inspection of the readiness and availability of risk-significant heat exchangers. The inspectors reviewed the data from a performance test for the division 1 emergency diesel generator diesel cooling water heat exchangers and verified the licensee used the industry standard periodic maintenance method outlined in EPRI NP-7552 for the heat exchanger. Additionally, the inspectors walked down the heat exchanger to observe its performance and material condition and verified that the heat exchanger was correctly categorized under the Maintenance Rule and was receiving the required maintenance.
These activities constitute completion of one heat sink performance annual review sample, as defined in Inspection Procedure 71111.07.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
.1 Annual Inspection
Inspection Scope
The inspector reviewed the annual operating examination test results for 2013. Since this was the first half of the biennial requalification cycle, the licensee was not required to administer a written examination. These results were assessed to determine if they were consistent with NUREG 1021, "Operator Licensing Examination Standards for Power Reactors," guidance and Inspection Manual Chapter 0609, Appendix I, "Licensed Operator Requalification Significance Determination Process," requirements. This review included the test results for a total of 7 crews composed of 57 licensed operators, which included 31 senior operators and 26 reactor operators. One crew and one individual failed the simulator portion of the exam. In addition, two individuals failed the Job Performance Measures (JPM) portion of the examination. Following the examination failures, the licensee remediated the affected individuals, and the individuals and crew passed a re-examination prior to resuming watch standing.
Findings
No findings were identified.
.2 Review of Licensed Operator Requalification
a. Inspection Scope
On October 9, 2013, 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 constitute completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
.3 Review of Licensed Operator Performance
a. Inspection Scope
On December 3, 2013, 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 Instrumentation and Control surveillances, troubleshooting of increased oxygen levels in the containment wetwell and removal from service of the reactor building intermediate range stack monitor for planned maintenance. The inspectors observed the operators performance during the following activities:
Shifting of tower makeup pumps to support vibration analysis
Authorization of work within the reactor building HVAC enclosure which renders secondary containment inoperable
Procedure OSP-INST-H101, Shift and Daily Instrument Checks (Modes 1, 2 and 3), Revision 80
In addition, the inspectors assessed the operators adherence to plant procedures, including OI-09, Operations Standards and Expectations and other operations department policies.
These activities constitute completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed the following three instances of degraded performance or condition of safety-related structures, systems, and components (SSCs):
September 27, 2013, Action Request AR 294644294644documenting a disconnected air line to the governor booster for the Division 1 emergency diesel generator
October 12, 2013, Action Request AR 292176292176documenting service water pump SW-P-1B entering the alert range during in-service testing
December 20, 2013, system review of containment monitoring 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 each 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.
These activities constituted completion of three maintenance effectiveness samples, as defined in Inspection Procedure 71111.12.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors reviewed the following three risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk:
September 24-27, 2013, yellow risk due to planned maintenance on emergency diesel generator 1
October 12, 2013, orange risk due to planned impeller lift adjustment on service water pump SW-P-1B
November 12, 2013, yellow risk due to planned maintenance on the reactor core isolation cooling system
The inspectors verified that these risk assessment were 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 assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments.
These activities constitute completion of three 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 structures, systems, or components (SSCs):
September 25, 2013, Action Request AR 294456294456 operability determination associated with lost position indication for control rod 26-15
November 6, 2013, Action Request AR 297128297128 operability determination associated with low service water flow and high temperature in analyzer room B
December 31, 2013, Action Request AR 300123300123 operability determination associated with missing tape wrap on an electrical connection for diesel generator 3
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. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability of the degraded SSC.
These activities constitute completion of three operability and functionality 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 structures, systems, or components (SSCs):
October 12, 2013, service water pump 1B following impeller lift adjustment
October 23, 2013, low pressure core spray pump room cooler RRA-CC-5 following planned cooling coil replacement
December 17, 2013, 250V DC battery E-B2-1 following repairs to high resistance connection
December 30, 2013, containment supply purge valve CSP-V-5 following relay replacement 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 constitute 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 structures, systems, and components (SSCs) were capable of performing their safety functions:
Other surveillance tests:
September 24, 2013, Procedure OSP-LPCS/IST-Q702, LPCS System Operability Test, Revision 34
October 1, 2013, Procedure ISP-FDR/EDR-M401, Drywell Sump Flow Monitors
- CFT, Revision 7
November 25, 2013, Procedure OSP-CCH/IST-M701, Control Room Emergency Chiller System A Operability, Revision 36
December 2, 2013, Procedure OSP-WMA-B701, Control Room Ventilation System A Pressurization Flow Test, Revision 18
December 5, 2013, Procedure ISP-MS-Q908, ATWS-RPT-ARI Actuation Reactor Level 2 Channels B and D - CFT/CC, Revision 6
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 test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected SSCs following testing.
These activities constitute 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 III, Design Control, for the licensees failure to translate the results of calculation NE-02-02-01, Control Room Boundary Leakage Limitation into allowed breach specifications for the control room ventilation boundary. Consequently, the specification used by operators in procedure PPM 1.3.57, Barrier Impairment, Revision 29 for determining the operability of the control room envelope was non-conservative with respect to station calculations.
Description.
On November 25, 2013, the inspectors identified air leakage around the seal of door C-509, which connects the Division 1 HVAC room and the cable chase room in the radioactive waste building. At the time of discovery, the control room envelope had 0.74 square inches of available margin out of a total allowable breach size of 26.16 square inches. The maximum allowed breach size is a specification used to determine operability of the control room envelope in procedure PPM 1.3.57, Barrier Impairment, Revision 29. Engineering reviewed the condition of door C-509 and determined that the identified seal leakage exceeded the remaining allowable breach size for the main control room envelope. Control room operators declared the control room envelope inoperable and entered Technical Specification action statement 3.7.3.B.
As part of their extent-of-condition review, the licensee identified three additional doors credited as part of the control room boundary with seal leakage. Engineering determined that the cumulative breach associated with these four doors was 72.40 square inches, well in excess of the maximum allowable breach size. The licensee sealed the affected doors with aluminum tape and documented the door seal leakage in Action Requests AR 298467298467and AR 298493298493
On December 4, 2013, the inspectors reviewed the maximum allowable breach size specification used in procedure PPM 1.3.57 to determine operability of the control room envelope. The maximum allowable breach specification is a calculated value based on performance of the control room emergency filtration system. To determine the maximum allowable breach, the licensee compares surveillance data from procedure OSP-WMA-B701, Control Room Ventilation System A Pressurization Flow Test, Revision 10, to calculation NE-02-02-01, Control Room Boundary Leakage Limitation, Revision 1, Table 1. The inspectors compared the results of procedure OSP-WMA-B701 performed on March 10, 2010 to Table 1 of NE-02-02-01 and determined that the results of the pressurization flow test would yield a maximum allowable control room envelope breach size of 21.07 square inches. The value determined by the inspectors was thus more restrictive than the 26.16 square inches currently specified to control room operators. The licensees engineering staff reviewed the March 10, 2010 surveillance data and determined that they had referenced the wrong revision of Table 1 in calculation NE-02-02-01 when providing specifications to control room operators, and that their error had resulted in a non-conservative error of 5.09 square inches. The inspectors determined that the cause of this error was a lack of formality in the process engineering personnel used to review the surveillance data. The inspectors also noted that in November 2013, the licensee revised the model work order for the surveillance to describe in more detail engineerings review of the surveillance data.
Because of the already low margin available for the control room envelope boundary, the licensees failure to translate the correct revision of Table 1 in calculation NE-02-02-01 into specifications resulted in periods between March 10, 2010 and November 20, 2012, when control room operators had unknowingly exceeded the maximum allowable control room breach size.
Following discovery of the subject error, the licensee updated the maximum allowed control room breach size to 34.18 square inches, a value which reflected the most recent performance of procedure OSP-WMA-B701, Control Room Ventilation System A Pressurization Flow Test, Revision 18, as performed on November 20, 2013.
The inspectors noted that when engineering personnel updated this value, they used the process described in the model work order for the surveillance.
On December 13, 2013, the licensee performed a past-operability review and determined that the non-conservative error did result in periods where the maximum allowable breach size was exceeded. However, the error did not result in an inoperable control room envelope boundary because the licensee was able to gain additional margin from conservative assumptions about other control room ventilation boundary breaches in effect during that period. The licensees failure to translate the correct revision of calculation NE-02-02-01 into specifications for operators was documented in Action Request AR 298914298914
Analysis.
The failure to translate the results of calculation NE-02-02-01, Control Room Boundary Leakage Limitation into allowed leakage specifications for the control room ventilation boundary was a performance deficiency. This performance deficiency was more than minor because it affected the design control attribute of the Barrier Integrity cornerstone objective of providing reasonable assurance physical design barriers protect the public from radionuclide releases caused by accidents or events, in that the performance deficiency resulted in periods when the maximum allowable control room breach size in station calculations was exceeded. The inspectors screened the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power." Using IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined this finding is of very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for by the control room. The finding did not have a cross-cutting aspect because the performance deficiency occurred in early 2010 using a process that is different from the process that currently exists and was therefore not reflective of current performance.
Enforcement.
10 CFR Part 50 Appendix B, Criterion III requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which [Appendix B] applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, from March 10, 2010 through November 20, 2013, the licensee failed to translate applicable regulatory requirements and the design basis into specifications for the control room envelope boundary. Specifically, the licensee failed to translate the results of calculation NE-02-02-01, Control Room Boundary Leakage Limitation, Revision 1 into the control room breach specification used by control room operators in procedure PPM 1.3.57, Barrier Impairment, Revision 25-29 used to determine the operability of the control room envelope. Following discovery of this issue, the licensee restored compliance by updating the maximum allowed leakage specification with the most recent surveillance data performed on November 20, 2013. Because this violation was of very low safety significance (Green) and was entered into the licensees corrective action program as Action Request AR 298914298914 this violation is being treated as a non-cited violation, consistent with the Enforcement Policy: NCV 05000397/2013005-02, Non-Conservative Error in Control Room Boundary Breach Specification.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (IP 71114.04)
a. Inspection Scope
The NSIR headquarters staff performed an in-office review of the latest revisions of various Emergency Plan Implementing Procedures (EPIPs) and the Emergency Plan located under ADAMS accession number ML13262A048 and listed in the Attachment.
The licensee determined that in accordance with 10 CFR 50.54(q), the changes made in the revisions resulted in no reduction in the effectiveness of the Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, this revision is subject to future inspection. The specific documents reviewed during this inspection are listed in the Attachment.
b. Findings
No findings were identified.
RADIATION SAFETY
Cornerstones: Public Radiation Safety and Occupational Radiation Safety
2RS5 Radiation Monitoring Instrumentation
a. Inspection Scope
This area was inspected to verify the licensee is assuring the accuracy and operability of radiation monitoring instruments that are used to:
- (1) monitor areas, materials, and workers to ensure a radiologically safe work environment; and
- (2) detect and quantify radioactive process streams and effluent releases. The inspectors used the requirements in 10 CFR Part 20, the technical specifications, and the licensees procedures required by technical specifications as criteria for determining compliance.
During the inspection, the inspectors interviewed licensee personnel, performed walkdowns of various portions of the plant, and reviewed the following items:
Selected plant configurations and alignments of process, postaccident, and effluent monitors with descriptions in the Final Safety Analysis Report and the offsite dose calculation manual
Select instrumentation, including effluent monitoring instrument, portable survey instruments, area radiation monitors, continuous air monitors, personnel contamination monitors, portal monitors, and small article monitors to examine their configurations and source checks
Calibration and testing of process and effluent monitors, laboratory instrumentation, whole body counters, postaccident monitoring instrumentation, portal monitors, personnel contamination monitors, small article monitors, portable survey instruments, area radiation monitors, electronic dosimetry, air samplers, continuous air monitors
Audits, self-assessments, and corrective action documents related to radiation monitoring instrumentation since the last inspection
Specific documents reviewed during this inspection are listed in the attachment.
These activities constitute completion of one sample as defined in Inspection Procedure 71124.05.
b. Findings
No findings were identified.
2RS6 Radioactive Gaseous and Liquid Effluent Treatment
a. Inspection Scope
This area was inspected to:
- (1) ensure the gaseous and liquid effluent processing systems are maintained so radiological discharges are properly mitigated, monitored, and evaluated with respect to public exposure;
- (2) ensure abnormal radioactive gaseous or liquid discharges and conditions, when effluent radiation monitors are out-of-service, are controlled in accordance with the applicable regulatory requirements and licensee procedures;
- (3) verify the licensees quality control program ensures the radioactive effluent sampling and analysis requirements are satisfied so discharges of radioactive materials are adequately quantified and evaluated; and
- (4) verify the adequacy of public dose projections resulting from radioactive effluent discharges. The inspectors used the requirements in 10 CFR Part 20; 10 CFR Part 50, Appendices A and I; 40 CFR Part 190; the Offsite Dose Calculation Manual, and licensee procedures required by the Technical Specifications as criteria for determining compliance. The inspectors interviewed licensee personnel and reviewed and/or observed the following items:
Radiological effluent release reports since the previous inspection and reports related to the effluent program issued since the previous inspection, if any
Effluent program implementing procedures, including sampling, monitor setpoint determinations and dose calculations
Equipment configuration and flow paths of selected gaseous and liquid discharge system components, filtered ventilation system material condition, and significant changes to their effluent release points, if any, and associated 10 CFR 50.59 reviews
Selected portions of the routine processing and discharge of radioactive gaseous and liquid effluents (including sample collection and analysis)
Controls used to ensure representative sampling and appropriate compensatory sampling
Results of the inter-laboratory comparison program
Effluent stack flow rates
Surveillance test results of technical specification-required ventilation effluent discharge systems since the previous inspection
Significant changes in reported dose values, if any
A selection of radioactive liquid and gaseous waste discharge permits Part 61 analyses and methods used to determine which isotopes are included in the source term
Offsite dose calculation manual changes, if any
Meteorological dispersion and deposition factors
Latest land use census
Records of abnormal gaseous or liquid tank discharges, if any
Groundwater monitoring results
Changes to the licensees written program for identifying and controlling contaminated spills/leaks to groundwater, if any
Identified leakage or spill events and entries made into 10 CFR 50.75
- (g) records, if any, and associated evaluations of the extent of the contamination and the radiological source term
Offsite notifications and reports of events associated with spills, leaks, or groundwater monitoring results, if any
Audits, self-assessments, reports, and corrective action documents related to radioactive gaseous and liquid effluent treatment since the last inspection
Specific documents reviewed during this inspection are listed in the Attachment.
These activities constitute completion of one sample, as defined in Inspection Procedure 71124.06.
b. Findings
No findings were identified.
2RS7 Radiological Environmental Monitoring Program
a. Inspection Scope
This area was inspected to:
- (1) ensure that the radiological environmental monitoring program verifies the impact of radioactive effluent releases to the environment and sufficiently validates the integrity of the radioactive gaseous and liquid effluent release program;
- (2) verify that the radiological environmental monitoring program is implemented consistent with the licensees technical specifications and/or offsite dose calculation manual, and to validate that the radioactive effluent release program meets the design objective contained in Appendix I to 10 CFR Part 50; and
- (3) ensure that the radiological environmental monitoring program monitors non-effluent exposure pathways, is based on sound principles and assumptions, and validates that doses to members of the public are within the dose limits of 10 CFR Part 20 and 40 CFR Part 190, as applicable. The inspectors reviewed and/or observed the following items:
Annual environmental monitoring reports and offsite dose calculation manual
Selected air sampling and thermoluminescence dosimeter monitoring stations
Collection and preparation of environmental samples
Operability, calibration, and maintenance of meteorological instruments
Selected events documented in the annual environmental monitoring report which involved a missed sample, inoperable sampler, lost thermoluminescence dosimeter, or anomalous measurement
Selected structures, systems, or components that may contain licensed material and has a credible mechanism for licensed material to reach ground water
Records required by 10 CFR 50.75(g)
Significant changes made by the licensee to the offsite dose calculation manual as the result of changes to the land census or sampler station modifications since the last inspection
Calibration and maintenance records for selected air samplers, composite water samplers, and environmental sample radiation measurement instrumentation
Interlaboratory comparison program results
Audits, self-assessments, reports, and corrective action documents related to the radiological environmental monitoring program since the last inspection
Specific documents reviewed during this inspection are listed in the Attachment.
These activities constitute completion of one sample 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
This area was inspected to verify the effectiveness of the licensees programs for processing, handling, storage, and transportation of radioactive material. The inspectors used the requirements of 10 CFR Parts 20, 61, and 71 and Department of Transportation regulations contained in 49 CFR Parts 171-180 for determining compliance. The inspectors interviewed licensee personnel and reviewed the following items:
The solid radioactive waste system description, process control program, and the scope of the licensees audit program
Control of radioactive waste storage areas including container labeling/marking and monitoring containers for deformation or signs of waste decomposition
Changes to the liquid and solid waste processing system configuration including a review of waste processing equipment that is not operational or abandoned in place
Radio-chemical sample analysis results for radioactive waste streams and use of scaling factors and calculations to account for difficult-to-measure radionuclides
Processes for waste classification including use of scaling factors and 10 CFR Part 61 analysis
Shipment packaging, surveying, labeling, marking, placarding, vehicle checking, driver instructing, and preparation of the disposal manifest
Audits, self-assessments, reports, and corrective action reports radioactive solid waste processing, and radioactive material handling, storage, and transportation performed since the last inspection
Specific documents reviewed during this inspection are listed in the Attachment.
These activities constitute completion of one sample 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 Mitigating Systems Performance Index: Heat Removal Systems (MS08)
a. Inspection Scope
The inspectors reviewed the licensees mitigating system performance index data for the period of October 2012 through September 2013 to verify the accuracy and completeness of the reported data. To determine the accuracy of the reported data, the inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6.
These activities constituted verification of the mitigating system performance index for heat removal systems, as defined in Inspection Procedure 71151.
b. Findings
No findings were identified.
.2 Mitigating Systems Performance Index: Cooling Water Support Systems (MS10)
a. Inspection Scope
The inspectors reviewed the licensees mitigating system performance index data for the period of October 2012 through September 2013 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 6, to determine the accuracy of the reported data.
These activities constituted verification of the mitigating system performance index for cooling water support systems, 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, 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 constitute completion of one semiannual trend review sample, as defined in Inspection Procedure 71152.
b. Findings
No findings were identified.
.3 Annual Follow-up of Selected Issues
a. Inspection Scope
The inspectors selected the following issue for an in-depth follow-up:
Action Request AR 292358292358documenting seat leakage past reactor core isolation cooling drain valves RCIC-V-25 and RCIC-V-26
The inspectors assessed the licensees problem identification threshold, cause analyses, extent of condition reviews and compensatory actions. The inspectors verified that the licensee appropriately prioritized the planned corrective actions and that these actions were adequate to correct the condition.
These activities constitute completion of one annual follow-up sample as defined in Inspection Procedure 71152.
b. Findings
No findings were identified.
4OA3 Follow-up of Events and Notices of Enforcement Discretion
.1 (Closed) Licensee Event Report 05000397/2013-004-01, Jumper makes Suppression
Pool Spray Valve Remote Transfer Switch Inoperable
On June 4, 2013, during performance of Procedure OSP-INST-B701, Remote Shutdown Panel Operability, Revision 17, the licensee discovered a factory-installed jumper across the remote transfer switch for valve RHR-V-27B. Subsequent review revealed that the jumper was left in place during motor control center replacement activities conducted in May 2011. Because the jumper made the remote transfer switch for valve RHR-V-27B inoperable for a period greater than allowed by Technical Specifications, the event was reportable under 10 CFR 50.73(a)(2)(i)(B). Inspectors reviewed the licensee event report associated with this event and determined that the report adequately documented the summary of the event including the cause of the event and potential safety consequences. The NRC previously identified non-cited violation 05000397/2013004-01, Improperly Installed Jumper Results in Inoperable Remote Transfer Switch documenting a violation of the plants technical specifications due to the improperly installed jumper. No additional performance deficiencies were identified. This licensee event report is closed.
.2 (Closed) Licensee Event Report 05000397/2013-006-00, Accidental Switch Bump
Makes High Pressure Core Spray and Diesel Inoperable
On June 27, 2013, a plant laborer was exiting the Division 3 diesel generator room when he inadvertently brushed against the control switch for the diesel mixed air fan DMA-FN-32 causing it to turn to the OFF position. The licensee declared the Division 3 emergency diesel generator, the high pressure core spray system, the Division 3 125 VDC battery charger, Division 3 battery, and the Division 3 AC electrical power distribution system inoperable due to the loss of power to its required support ventilation system. In response to the loss of power which was annunciated in the main control room, plant operators restored the fan control switch back to ON position approximately 21 minutes after initially bumped by the plant laborer. The loss of the high pressure core spray system resulted in the temporary loss of safety function for a single train system which is reportable under 10 CFR 50.73(a)(2)(v). The inspectors reviewed the licensee event report associated with this event and determined that the report adequately documented the summary of the event including the cause of the event and potential safety consequences. No performance deficiencies were identified. This licensee event report is closed.
These activities constitute completion of two event follow-up samples, as defined in Inspection Procedure 71153.
4OA5 Other Activities
Inspection Procedure 92723 - Follow Up Inspection for Three or More Severity Level IV Traditional Enforcement Violations in the Same Area in a 12-Month Period
a. Inspection Scope
The inspectors performed Inspection Procedure 92723, Follow Up Inspection for Three or More Severity Level IV Traditional Enforcement Violations in the Same Area in a 12-Month Period, in accordance with the assessment letter dated September 3, 2013 (ML13246A431). Specifically, the inspectors examined the licensees response to four violations associated with impeding the regulatory process. These violations involved the following regulatory issues:
Failure to ensure integrity of licensed operator examinations (10 CFR 55.49)
Failure to obtain a license amendment prior to implementing a change to the Main Control Room heating, ventilation, and air conditioning system. (10 CFR 50.59)
Failure to obtain a license amendment prior to implementing a change to the Reactor Water Cleanup system piping. (10 CFR 50.59)
Failure to update the Final Safety Analysis Report and implement commitments to inspect the Diesel Fuel Oil Storage Tanks (10 CFR 50.71e)
Documents reviewed by the inspectors are listed in the attachment.
b. Findings and Observations
Based on the review of the licensees apparent cause evaluation, as documented in Action Request AR 293188293188 the inspectors determined that an adequate assessment of these conditions had been performed including the identification of common causes. The inspectors also determined that the licensee had implemented a sufficient range of corrective actions to address the identified common causes and minimize the potential for reoccurrence of these issues.
No findings were identified.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On November 7, 2013, the inspectors presented the results of the radiation safety inspections to Mr. J. Dittmer, Acting Vice President, Engineering, and other members of the licensee staff. On November 26, 2013, the inspectors conducted a teleconference with Mr. Grover Hettel, Vice President, Operations, and other members of the licensee staff to discuss the re-characterization of findings associated with Sections 2RS5 and 2RS6 of this report. The licensee acknowledged the issues presented.
On December 26, 2013, the inspector presented the results of the licensed operator annual requalification examination results review to Mr. Ron Hayden, Exam Developer. The licensee acknowledged the results of the inspection.
On January 6, 2014, the inspectors presented the inspection results to Mr. M. Reddemann, Chief Executive Officer, and other members of the licensee staff. The licensee acknowledged the issues presented.
The inspectors confirmed that proprietary information was not provided or examined during the inspections.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation.
.1 Technical Specification 5.4.1.a, requires, in part, that written procedures be established,
implemented, and maintained as described in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Regulatory Guide 1.33, Paragraph 9.a states that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, the licensee failed to preplan written procedures appropriate to the circumstances as required by Technical Specification 5.4.1.a. Specifically, on April 25, 2011 and January 4, 2010, the licensee planned Work Orders 01182801 and 01179746 to adjust the impeller lift for standby service water pumps 1A and 1B but failed to specify the appropriate impeller lift for the pump. Consequently, the licensee set the impeller lift higher than specified in the pump design during performance of Work Orders 01182801 and 01179746, resulting in degraded performance of the pump. The licensee identified this violation and entered it into their corrective action program as Action Requests AR 219553219553and AR 292521292521 This finding was determined to be of very low safety significance because the finding is a design or qualification deficiency that did not affect operability or functionality.
.2 Title 10 CFR 71.5(a) requires that each licensee who transports material outside the
site of usage, as specified in the NRC licenseshall comply with the applicable requirements of the Department of Transportation regulations in 49 CFR
Title 49 CFR 172.203(d)(3) states, in part, the description for a shipment of a Class 7 (radioactive) material must include the activity contained in each package of the shipment. Contrary to the above, on August 29, 2011, in the shipping manifest for Radioactive Waste Shipment 11-106, the licensee did not include the correct activity for the radioactive mechanical filters shipped, in that the original manifest stated an activity of 8.09 Curies, but the licensee later determined the activity was 31.54 Curies. The licensee identified this violation and entered it into their corrective action program as Action Request AR 248151248151 This finding was determined to be of very low safety significance (Green) because the finding did not involve the radioactive effluent release program or the radiological environmental monitoring program, but the finding did involve the transportation of radioactive material. The finding was not
- (1) in excess of radiation limits,
- (2) a breach of package during transit,
- (3) a certificate of compliance issue,
- (4) a low-level burial ground noncompliance, or
- (5) a failure to make notifications or provide emergency information.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- P. Allen, System Engineer, System Engineering
- J. Brown, Operations Supervisor, Chemistry
- G. Burton, Health Physics Staff Advisor, Radiological Support
- N. Chada, Engineer Senior, System Engineering
- K. Clark, Environmental Scientist II, Environmental Services
- J. Darling, NSSS Supervisor, System Engineering
- D. Gregoire, Regulatory Affairs Manager
- R. Hayden, Exam Developer
- M. Hedges, Principle Licensing Engineer, Regulatory Affairs
- M. Holle, System Engineer, System Engineering
- C. Hove, Technician, Radiation Protection
- M. Kinmark, Health Physics Staff Advisor, Radiological Support
- M. Laudisio, Supervisor, Radiological Support
- C. Madden, Scientist IV, Chemistry
- D. Mee, Environmental Scientist III, Environmental Services
- T. Northstrom, Supervisor, Environmental Services
- S. Nappi, Corrective Action Program Supervisor
- E. Nguyen, Environmental Scientist II, Environmental Services
- J. Pierce, Manager, Chemistry
- A. Pierson, Specialist III, Chemistry
- R. Sanker, Supervisor, Radiological Operations
- R. Schott, Technician, Chemistry Support
- B. Schuetz, Maintenance Manager
- M. Shobe, Technician, Chemistry
- D. Suarez, Licensing Engineer, Regulatory Affairs
- J. Tarr, Rad Waste Technician, Radiation Protection
- J. Trautvetter, Compliance Supervisor, Regulatory Affairs
- K. Webb, Technician, Radiation Protection
- R. Williams, Rad Waste Technician, Radiation Protection
- D. Wolfgramm, Licensing Engineer, Regulatory Affairs
NRC Personnel
- V. Gaddy, Branch Chief
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000397-2013005-01 NCV Failure to Translate Internal Flooding Design into Station Procedures (Section 1R06)
- 05000397-2013005-02 NCV Non-Conservative Error in Control Room Boundary Breach Specification (Section 1R22)
Closed
- 05000397-2013-004-01 LER Jumper makes Suppression Pool Spray Valve Remote Transfer Switch Inoperable
- 05000397-2013-006-00 LER Accidental Switch Bump Makes High Pressure Core Spray and Diesel Inoperable