IR 05000445/1993033
| ML20059D205 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 10/25/1993 |
| From: | Yandell L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20059D168 | List: |
| References | |
| 50-445-93-33, 50-446-93-33, NUDOCS 9311020162 | |
| Download: ML20059D205 (21) | |
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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION '
REGION IV
Inspection Report: 50-445/93-33 50-446/93-33 Licenses: NPF-87 ,
NPF-89 ,
Licensee: TV Electric Skyway Tower 400 North Olive Street, L.B. 81 :
Dallas, Texas s
facility Name: Comanche Peak Steam Electric Station, Units 1 and 2 Inspection At: Glen Rose, Texas Inspection Conducted: August 22 through October 2, 1993 Inspectors: D. N. Graves, Senior Resident Inspector G. E. Werner, Resident Inspector -
K. M. Kennedy, Resident Inspector Approved: u bbh
'Eate-28 1993
L. A. ~Ya6 dell, Ch'ief, Project Section B ,
Division of Reactor Projects Inspection Summary Areas Inspected (Units 1 and 2): Routine, unannounced inspection of operational safety verification; maintenance and surveillance observations; Temporary Instruction 2500/028, " Employee Concern Program"; followup on corrective actions for violations; and other followu Results (Units 1 and 2):
- A drain line between a Unit 2 feedwater isolation valve and the containment was found uncapped by the licensee. Investigation by the licensee revealed that the system operating procedures did not reflect-the requirements of the Final Safety Analysis Review (FSAR) and the design basis documents regarding caps. This was identified as a noncited violation (Section 2.2).
- An excellent prejob briefing was conducted by radiation protection prior l to a Unit 2 containment entry (Section 2.4).
- A minor weakness in component turnover from construction to operations- ,
was identified. Cloth foam filter material, remaining from.the l
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9311020162 931027 PDR ADOCK 05000445 'l f
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-2-preoperational period, was found covering the air intakes of the motors on Centrifugal Charging Pump 2-01 and Positive Displacement Charging Pump 2-01 (Section 2.6).
- A security guard did not undergo a search for contraband following entry through the personnel monitors. A violation of station procedures would have resulted had he entered the protected area without undergoing a *
search (Section 2.8).
- Excellent coordination between operations and station nuclear / reactor engineering was observed during the Unit 2 reactor startup (Section 2.9).
- All maintenance and surveillance activities were well performed, with excellent self-verification practices utilized by the instrument and control technicians (Sections 3 and 4).
Summar_y of Inspection Findings:
A noncited violation was identified (Section 2.2).
- Temporary Instruction 2500/028, " Employee Concerns Program," was completed and closed (Section 5).
Violation 446/9326-01.was closed (Section 6).
- Inspection Followup Item 445/9214-03 was closed (Section 7).
Attachments:
- Attachment 1 - Persons Contacted and Exit Meeting
- Attachment 2 - Employee Concern Program Survey
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-3-DETAILS 1 PLANT STATUS Unit 1 operated at full power for the duration of this inspection perio Unit 2 operated at full power from the beginning of this inspection period until August 17 when the unit was shutdown via a normal reactor shutdown to perform repairs to a feedwater split flow bypass valve inside containmen The unit was restarted on September 29, with power ascension to 100 percent in progress at the end of this inspection perio OPERATIONAL SAFETY VERIFICATION (71707) ,
2.1 Plant Tours r
A review of a sample of roving fire watch logs indicated that the roving fire watch inspections were performed at the intervals specified in Procedure STA-729, Revision 4, " Control of Transient Combustibles, Ignition Sources and Fire Watches."
During a tour of the plant, inspectors identified that a tubular frame constructed around a leaking steam generator blowdown valve in the Unit 2 penetration valve room was freestanding and unsteady. An information tag attached to the fran.e indicated that the frame was erected to support a containment tent around a leaking valve although there was no containment on the frame. The inspectors were concerned about the stability of this frame and brought it to the attention of the control room operators, who had the frame remove General housekeeping was determined to be very goo .2 Uncapped Drain Valve On Containment Penetration Line On August 15, 1993, while performing a surveillance on containment nonautomatic isolation components, the licensee determined that Valve 2FW-Olll, a drain valve on a main feedwater header between the feedwater isolation valve and the containment penetration, was found to not have a cap installed as required by the surveillance. The licensee could not install a cap on the drain because it was leaking steam and a section of hose was installed on the drain routing the leakage into a floor drain. The Technical Specification 3.6.1.1 action statement regarding primary _ containment integrity:
was entered which required containment integrity be established within I hour or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A review of the mechanical drawing, M2-203, Sheet 1, and System Operating Procedure SOP-302,
"Feedwater System," by the licensee initially concluded that the drain did not need to be capped and the action statement was exited prior to the expiration of I hour from the time of identification. Operations Notification and Evaluation (ONE) Form 93-1548 was initiated, and subsequently upgraded to a plant incident report to document and resolve the apparent procedural discrepancies. A " quick turnaround" technical evaluation was performed as a
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I-4-result of the ONE form and identified additional information regarding the capping of vents and drains inside the boundary of containment isolation valves, as well as an evaluation of the observed leakage and its potential radiological consequences in the event of an acciden The engineering evaluation of the observed leakage concluded that the leakage, approximately 2 gallons per minute, posed no significant radiological risk and ,
that the consequences of an accident would have been much less than the design basis analysis. The technical evaluation also indicated that Note 26 on .
Drawing M1-200, which was applicable to all vital station drawings, FSAR Section 6.2.4.1.3, and Design Basis Document DBD-ME-013, Section 5.4, all required that local vent, drain, and test connections within the containment isolation boundary be capped to ensure containment integrit The licensee's investigation of the removal of the cap and connection of the hose could not determine the precise date of cap removal. The surveillance to *
verify that caps were installed was performed on June 23, 1993, and verified ,
that Valve 2FW-0111 was capped. A work request was written on July 9, identifying that Valve 2FW-Olll was leaking by and that the cap could not be -
installed due to the leak. The surveillance verifying capping was again performed on July 23 and verified that the cap was once again installed. No documentation of any work on the valve subsequent to the initiation of the '
work request could be located until the identification of the missing cap on August 15. Following the identification of the missing cap, the steam leak
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was stopped and the drain line was plugged by the injection of leak sealing material and was cappe In conclusion, the documentation was indeterminate >
with respect to adequacy of surveillance performance and record keepin As a result of the discrepancies between the system operating procedures, the station vital drawings, the surveillance procedure, and the requirements of the FSAR and the design basis document, a number of corrective actions were initiated by the licensee. The system operating procedures for all procedures containing vents, drains, or test connections within the boundary of a containment isolation valve were reviewed and revised to add the requirement for caps. Revisions will be initiated to the FSAR and the technical ,
requirements manual to clarify the basis for capping. The station vital drawings and the design basis document will be revised to remove or add. caps as needed to match the revised requirements and/or. acceptance criteria. A unique label or identification method for caps required to meet containment integrity will be implemente The inspector reviewed the plant incident report, including the root cause determination, and concluded that the licensee had performed a comprehensive investigation and evaluation of the even :
The safety significance of this particular occurrence was low due to the fact l that it occurred on a closed system. The corrective actions initiated and '
planned were comprehensive and should, when fully implemented, prevent recurrence. Although the failure of the system operating procedures to ,
accurately reflect the requirements of the design basis documents and the FSAR j I
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-5-is a violation of Technical Specification 6.8.1 and Regulatory Guide 1.33, Appendix A, Section 3.f.(1), it will not be cited as a result of the licensee's identification of the condition and prompt initiation of corrective action .3 Unit 1 New Fuel Receipt Inspection
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On August 24, 1993, the inspector observed activities associated with the receipt, inspection, and storage of new fuel supplied by Siemens Power Corporation. This activity was performed in accordance with Procedure RF0-201, Revision 7, " Receipt, Inspection and Storage of New Fuel-And Insert Core Components." The inspector noted that all observed activities were conducted in accordance with the procedure. A radiation protection technician was present during the evolution to perform contamination and radiation surveys and to enrura proper radiological precautions were take The fuel receipt inspector wet observed performing visual inspection of the fuel shipping container interior, the fuel assembly protective liner, and the fuel assembly. These inspections were thorough and conducted in accordance with the procedure.. A fuel handling supervisor supervised the movement of the fuel assembly from the shipping container to the new fuel inspection station ,
and then to the new fuel storage rack locatio ' procedural weakness was identified with Procedure RFO-201, Revision 7, in .
nat Step 6.8.1.9, which provided instructions for the proper orientation of a Westinghouse fuel assembly when placing it in the bottom nozzle support plate of the new fuel inspection station, was not applicable to fuel assemblies provided by Siemens Power Corporation. The fuel handling supervisor indicated that he was aware of the procedure deficiency and that the Siemens fuel assemblies were being placed in the inspection stand with the proper ,
orientation. This deficiency did not affect the proper orientation of the
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fuel assembly in the new fuel storage rack .4 Unit 2 Containment Entry ;
On August 28 the inspectors accompanied an auxiliary operator and the feedwater system engineer on the walkdown of Unit 2 containment. At this time, operations personnel were pumping approximately 1200 gallons per day of nonradioactive secondary water from the containment sumps. The licensee was attempting to identify any increased leakage from previously identified components and to quantify additional leakage paths and plan appropriate r corrective maintenance. During the walkdown, Chemical Feedwater Relief Valve 2CF-0067 was identified as a new leakage source and known leakage from ,
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Feedwater Split Flow Bypass Valve 2FV-2183 had increased in magnitude. These -
two valves contributed to almost all identified secondary leakag Radiation protection personnel provided an excellent prejob briefing that identified radiation, high radiation, and contamination areas. Previous survey maps were reviewed with the containment entry team, and access to areas that were prohibited were discussed. Continuous radiation protection coverage was provide ,
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-6-2.5 Unit 2 Containment Penetration Verification Prior to Mode 4 entry, the inspectors performed a walkdown of manual penetrations inside the reactor containment using Valve Lineup Sheet OPT-219B-1, " Containment Penetration Non-Automatic Isolation Component (IRC) Position Verification Data Sheet." All accessible valves were verified-to be in the required position. The inspectors identified several valves with '
heavy boron deposits on the valve stem and body area. These valves had been identified by the licensee and were being tracked as part of previous containment walkdown discrepancie .6 Unit E_Centrifuaal Chargina Pump Motor Vent Screens During a tour of the auxiliary building on August 30, 1993, the inspector observed that, Centrifugal Charging Pump 2-01 had a cloth foam filter and screen mater #.al covering the inlet and outlet motor air vents. The inspector did not observe this material on the other centrifugal charging pumps for either unit. The licensee evaluated this condition and determined that this material did not affect the operation of the pump motor. They believed that the material was installed during Unit 2 construction for cleanliness control and had not been removed prior to plant startup. The: licensee inspected all Units 1 and 2 safeguards pumps to identify similar conditions and, of the 26 pumps inspected, identified this condition on one additional pump, Positive Displacement Pump 2-01. The material was removed frcm Centrifugal Charging Pump 2-01 and was scheduled to be removed from Positive Displacement Pump 2-01 '
during the next available equipment outage period. These appear to be isolated instances of weaknesses in the system turnover proces .7 Unft 2 Shutdown Control Room Observations The inspectors observed portions of the Unit 2 reactor shutdown on September 17, 18, and 19. Good coordination of work activities and plant shutdown functions were accomplished by the interface between the extra unit supervisor and the on-duty unit supervisor. The unit supervisor's main duties were focused on plant shutdown and excellent oversight of plant equipment operation was maintaine Plant cooldown was monitored by the reactor operator on 15 minute intervals -j per Procedure OPT-407, "RCS Temperature and Pressure Verification,"
Revision 5. All cooldown temperature readings were well within Technical Specifications requirements, and were also well below the administrative limits of less than 60oF/Hr for the reactor coolant system and less than ,
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100oF/Hr for the pressurize All emergency core cooling equipment were verified in their correct lineups as specified by the Technical Specification i
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2.8 Security Observations On September 20, while entering the primary access point, the inspectors noted that a security guard who was assigned to the post had exited the primary access point and returned to his duties inside the primary access point .
without undergoing a search for weapons and explosives. The guard passed
through both electronic monitoring detectors which subsequently alarmed and he -
was not physically searched for contraband. This observations was discussed with security managemen Security Field Report 1856-93 was written to document the apparent security violation. Security management presented their findings and interpretation of the event to the inspector The managers indicated that their expectations were not followe As stated by the managers, if an on-duty security officer leaves the primary access point and is out of the line-of-sight of the other security officers, that officer is to undergo a search for firearms, explosives, and incendiary devices prior to entering the protected are *
These searches were required prior to entering the protected area by Security Procedure SEC-302, " Personnel Identification, Key Card Badge Issuance, and-Access Control," Revision 7; and Security Post Order (Post 8), " Primary Access Point Personnel, Package, Material Search Officer." Since the security officer never entered the protected area after this observation (verified by access computer), no violation occurred; however, this was identified to the licensee as a poor practice that could lead to a violation of security procedure .
Currently, the licensee is requiring all officers to be either electronically or physically searched whenever they enter the protected area or the search areas enclosed by the screen barrier with no exceptions (delineated-in '
TU Memorandums TSEC 93092 and BRNS 93946).
l 2.9 Unit 2 Reactor Startup On September 29, the inspector observed the Unit 2 reactor startup following
the maintenance outage to repair a leaking split-flow bypass valve flange in the main feedwater system. While performing a criticality prediction using an inverse count ratio, the station nuclear engineers predicted that criticality {
would occur at a control rod height below the -500 pcm tolerance. allowed per i the_ plant startup procedure. The startup was terminated and control rods were l reinserted such that Control Bank A was at 50 steps. The estimated critical !
condition was recalculated, verified correct, and TU Electric Reactor !
Engineering was notified. ONE Form 93-1737 was generated to document and ,
evaluate the condition. Reactor engineering, in conjunction with the station nuclear engineers, concluded that the discrepancy in the predicted critical condition utilizing the inverse count ratio was due to two factors. One, the calculated data did not take into account the buildup of plutonium:following shutdown, which would add positive reactivity estimated to be approximately 300 pc Second, the inverse count ratio data included a core configuration in which Control Bank C rods were being withdrawn in the vicinity of the primary neutron source. This resulted in a condition in which the observed d
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-8-t count rate increase was not representative of full core conditions and resulted in a greater increase in source range counts in the vicinity of the source range detector. Plotting of the inverse count ratio points indicated that criticality would occur earlier than origir. ally calculate Previous Unit 2 information from data taken with Control Bank C in a similar position overestimated the core reactivity by at least 100 pc Based on the above two factors, reactor engineering provided a technical evaluation recommending continuing the approach to critical utilizing an additional 300 pcm increase in core reactivity due to plutonium buildu The crew briefing prior to the initial startup attempt included a discussion of the possibility of plutonium buildup affecting the accuracy of the estimated critical position, and that the operators should be sensitive to that effec During the second startup attempt, the inverse count ratio predictions were within the adjusted reactivity range, and the actual critical rod position was within the -500 pcm value from the original estimated critical positio .10 Unit 1 Outage Reviews The inspectors conducted a review of various Unit 1 outage items in preparation for Refueling Outage 3 (1RF03). Those items evaluated included additional radiological worker training for workers and supervisors to address poor radiological work practices identified by both the licensee and the NRC (documented in Inspection Reports 50-445/92-47; 50-446/92-47 and 50-445/92-59; 50-446/92-59), all Unit 1 maintenance items which require an outage condition but are not scheduled for 1RF03, Unit 1 Refueling Outage 2 (IRF02) critique summaries and their dispositions, and the Independent Safety Engineering ,
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Group's (ISEG) field notes concerning the licensee's preparation for 1RF0 Additional radiological training was provided to TV' Electric supervisors, contractor personnel, and to some TU Electric craft workers. The supervisors >
were given training on observation guidelines for work activities inside the radiological controlled areas. All the lesson plans discussed Unit I refueling outage lessons learned, examples of poor radiological work practices, and individual responsibilities. The inspectors-reviewed the training plans and found them.to cover material necessary to address previous weaknesses. Additionally, a training roster-for the: supervisory training was ,
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reviewed and the majority of the personnel scheduled to attend were'presen However, even though operations personnel exhibited similar poor radiological work practices during 1RF02 as documented in NRC Inspection Report 50-445/92-59; 50-446/92-59, only 14 percent of the operations supervisors attended the enhanced radiation protection observational training session Approximately 140 1RF03 deferred Unit 1 outage items were reviewed as well as 961RF02 critique summaries and 25 ISEG field notes. No safety significant *
problems were identified during the reviews; however, ISEG Field Note 93-493 i
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did identify six cases where the operations department staffing could be insufficient to handle the large number of clearances scheduled. These I activities will be observed during the normal outage inspection activitie The inspectors also observed a management briefing regarding risk management presented by a member of the ISEG. The briefing focused on the importance of assessing changes in schedule or scope of work activity and how those changes -
affect the outage risk assessment. The briefing also highlighted a document to be prepared daily during the outage reflecting the time to core boiling and the operability of core cooling systems. Overall, the briefing was well conducted, informative, and appeared to be well received by all personnel in attendanc .11 Conclusions ;
The licensee was found to be operating the plant safely and generally in ,
accordance with established procedures. Excellent coordination between operations and the station nuclear engineers was observed during the Unit 2 reactor startup. Although several deficiencies were noted by the inspectors, ,
they were of minor safety significance and promptly corrected by the license The system operating procedures regarding the capping of vents and drains within the boundaries of containment isolation valves were examples of inadequate procedures in that they did not reflect the requirements of the FSAR and the design basis document and resulted in a noncited violatio MAINTENANCE OBSERVATIONS (62703)
3.1 Unit 2 Source Range Instrumentation Troubleshootinq
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l Source Range Instrument 2N-0031 failed to energize during the shutdown and subsequently spiked during initial troubleshooting causing a flux doubling i actuation. The reactor protection actuation was documented on a ONE form and l
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reported per 10 CFR 50.72(b)(',(ii). Instrumentation and control (I&C)
technicians were observed troubleshooting Source Range Instrument 2N-0031 the I following day under Work Order 1-93-053110-00. The system engineer assisted the technicians during the maintenance activity. Technical Manual CP-0001-059, " Nuclear Instrumentation System," was used to troubleshoot the faulted channel; however, no reason for the chantiel failure was found and the source range instrument was later returned to service. Good work j practices were used by the two technicians, !
3.2 Steam Generator (SG) 2-01 Inspection Cover Maintenance During the initial walkdown of Unit 2 containment following the maintenance shutdown, the licensee identified a feedwater leak at a 6-inch SG inspection cover. Mechanical maintenance personnel performed the inspection, repair, and reassembly of the inspection cover under Work Order 1-93-054725-00. The flex gasket was found to have failed with no damage to the flex seating surfaces on i
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either the SG flange opening or the cover. A system engineer assisted the mechanics in determining the appropriate rework scope which entailed only a
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q flex gasket replacement. The mechanics used good work practices by ensuring that all sealing surfaces were properly cleaned and inspected for flatness. A radiation protection technician provided coverage during the maintenance ,
i activit .3 Diesel Generator (DG) 1-01 Firino Pressures ,
Mechanical maintenance technicians were observed obtaining cold firing -
pressures for DG 1-01. The technicians were performing the activity under Work Order 3-93-301337-01 and Procedure MSM-P0-3359, " Emergency Diesel Firing Pressure," Revision 2. A special precaution included in the work order required the technicians to wear face protection while obtaining firing pressures; however, the technicians only had on safety glasses. When .
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questioned, the technicians felt they only needed safety glasses and did not obtain a face shield. The mechanical maintenance planners were investigating this requirement to determine if it was necessary for future task Data was being collected prior to 1RF03 in order to determine the condition of l the DG's valves and compression rings. Both the licensee and vendor will evaluate the cold firing pressures and plan corrective maintenance, if neede All firing pressure.s were within 10 percent of each other. The system engi.._ r was present and indicated that the initial data appeared to be ;
acceptaul The inspectors reviewed the DG's operating parameters and found that most parameters were within the required specifications. Those parameters found out-of-specification were appropriately annotated in the auxiliary operator's logs as being attributed to the DG being run in an unloaded conditio .4 Repair of Unit 1 New Fuel Assembly During a new fuel assembly receipt inspection conducted on August 31, 1993, ,
the licensee discovered that the screws used to attach the lower tie plate to the guide tubes were not crimped on Fuel Assembly F01. The vendor informed the licensee that the lower tie plates on 18 Comanche Peak fuel. assemblies recently shipped to the site had been replaced at the fuel fabrication facility, including Fuel Assembly F01. The other 17 assemblies had recently undergone new fuel receipt inspections and had been placed in the new fuel storage racks. After the condition on Fuel Assembly F01 was identified, the licensee reinspected these 17 fuel assemblies and did not find a similar proble .
On September 3, 1993, the inspectors observed the licensee perform repairs on Fuel Assembly F01. The repairs were performed by vendor representatives and inspected by a licensee quality control inspector. The inspectors determined that the repairs were performed in accordance with the procedure, that the correct tools were used, and that the equipment requiring calibration was current. Proper cleanliness controls were.being implemented in the ma of the repair activity; however, personnel were observed entering and ex aing the '
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area above the new fuel storage racks without logging in or out in the personnel accountability lo .5 Conclusions i
All of the observed maintenance activities were well conducted with the technicians displaying good work practices. One instance of a technician not '
wearing the required face protection was observed, and that requirement was being reviewed by the licensee as to its necessit SURVEILLANCE OBSERVATIONS (61726)
4.1 Unit 2 Train B K610 Slave Relay Test
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Slave Relay Test Procedure OPT-492B, Revision 1, " Train B Safeguards Slave Relay K610 Actuation Test," was conducted to verify operability of sections of the engineered safety feature actuation system instrumentation (Work Order 5-93-502947-AC). The licensee conducted an adequate pre-evolution brief in accordance with Procedure ODA-407, Revision 4, " Guideline on Use of Procedures," Form 4; " Checklist for High Risk, Infrequent Evolutions or -
Heighten Level Awareness Activities."
Two reactor operators conducted the test with intermittent supervision supplied by the unit supervisor. One reactor operator was the reader while the other performed the required steps. Both reactor operators used excellent- i self-checking and communications. Outstanding verbalization concerning required manipulations and expected responses by the reader allowed the reactor operators performing the steps to precisely understand and execute each step. The surveillance was completed with all acceptance criteria ,
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4.2 Unit 2 Train B K601 Slave Relay Test The inspectors observed a reactor operators performing Surveillance Procedure OPT-487B, " Train B Safeguards Slave Relay _K601 Actuation Test,"'
Revision 1 (Work Order 5-93-502939-AE). This test caused various safety-related pumps to start, heating and ventilation systems realignments, loss of several motor control centers, and actuation of Train B safety injection sequence The reactor operators ensured that loads that were lost would not effect required equipment and made several plant announcements concerning the loss of portions of containment lighting and the elevator since personnel were working in containment. Good preparation and coordination by control room personnel allowed the test to be completed satisfactorily with no unexpected equipment actuations or realignment l F
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4.3 Unit 1 SG Level Channel 0517 Analog Channel Operational Test t
I&C technicians were observed performing an analog channel operational test ,
for SG narrow range level Channel 0517 in accordance with Procedure INC-7322A, ,
" Analog Channel Operational Test and Channel Calibration Steam Generator NR Level, Loop 1, Protection Set IV Channel 0517," Revision 4; and Work ,
Order 5-93-501775-AC. The test checked the accuracies of the channel sensor- -
and would have allowed for calibration if the as-found values were unacceptable. All as-found values were well within the acceptable range and no recalibration of the channel was necessary. The inspectors verified that both digital multiple meters used to collect voltage data were within their-respective calibration due date The technician performing the surveillance had previously completed this surveillance; however, he indicated it had been a considerable time since his last performance. Additionally, a trainee was assisting in the test. Both '
technicians practiced excellent self-verification techniques and proper two-way communication. The lead technician allowed the trainee to perform the surveillance while constantly monitoring the individual. The pace was slow and deliberate thereby allowing the lead technician to verify the trainee's switch and test lead positioning before proceedin .4 Unit 2 Solid State Protection System Actuation Logic Test On August 27, 1993, the inspector observed the performance of a surveillance test conducted in accordance with Procedure OPT-448B, Revision 0, " Mode 1, 3 and 4 Solid State Protection System, Train 'B' Actuation . Logic Test."
Prior to the start of the test a crew briefing was conducted by the unit supervisor with those personnel involved in the performance of the surveillance test. In addition to using briefing notes included with the surveillance procedure, the unit supervisor utilized the " Checklist For~High Risk, Infrequent Evolutions Or Heighten Level Of Awareness Activities," found in Procedure ODA-407, Revision 4, " Guideline On Use Of Procedures." The pre-evolution brief was thorough and covered the objectives contained in these two checklist As recommended in Procedure OPT-4488, two operators performed the surveillance: one operator read the procedural steps and the other operator performed the steps. In addition, the system engineer was present during the conduct of the test. The surveillance was performed in accordance with the ,
test procedure and all acceptance criteria were satisfied. No deficiencies were noted during the performance of the surveillanc .5 Ionization Detector Test On September 15, 1993, the inspector observed technicians perform surveillances on several ionization detectors associated with safeguards building Elevation 810, Train A Switchgear Room Fire Protection Panel 1-LV-33A. This surveillance was conducted in accoraance with
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-13-Procedure MSE-CO-7335, Revision 1, " Ionization Detector Sensitivity Test and Visual Check." The inspector determined that the technician was knowledgeable, performed the surveillances in accordance with the procedure, and that the calibration of the measuring and test equipment was curren .6 Train B Safeguards Slave Relay K624 Actuation Test
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On September 21, 1993, the inspector observed control room operators perform Procedure OPT-4868, Revision 1, " Train B Safeguards Slave Relay K624 Actuation Test." No discrepancies were note .7 Unit 1 Slave Relay K645 Test On September 14, 1993, the inspector observed the performance of the testing of Slave Relay K645B and the subsequent test of Containment Spray Pump 1-02 which was started by actuation of the slave relay. Both tests were performed in accordance with their respective surveillance procedures and were properly authorized and executed. Initial test results indicated satisfactory performance and the surveillance acceptance criteria were in conformance with the requirements of the Technical Specifications and the FSA .8 Conclusions The inspectors observed that all of the witnessed surveillance activities were -
well performed. The acceptance criteria, where reviewed, were appropriate for the surveillance. Excellent self-verification practices were utilized by the instrument and control technicians during performance of the test TEMPORARY INSTRUCTION 2500/028, " EMPLOYEE CONCERNS PROGRAM" (2500/02.8)
This temperary instruction was to determine the characteristics of employee concern programs that licensees have implemented to provide their employees, who wish to raise safety issues, an alternate path from their supervisor or normal line management to express these concerns and to encourage people to come forward with their concerns without fear of retributio .1 Discussion The inspector reviewed the employee concerns program known as SAFETEAM and met ,
with the Manager, SAFETEAM on August 17, 1993, to discuss the program as implemented by TV Electric and to collect the information requested by the temporary instructio .2 Conclusions I
Information on the program was summarized and is included in this inspection report as Attachment t
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-14-6 FOLLOWUP ON CORRECTIVE ACTIONS FOR VIOLATIONS (92702)
IClosed) Violation 50-446/9326-01: Unit 2 Motor-Driven Auxiliary feedwater (MDAFW) Pump 2-01 Packing Replacement Desian Change Authorization (DCA)
This violation involved the licensee's failure to perform an engineering review and justification for a change to a DCA for MDAFW Pump 2-01 packing -
configuration. On June 16, 1993, the inspectors identified that the installation of the packing gland follower for MDAFW Pump 2-10 was not in accordance with the DCA. The pump was released for operation by a mechanical maintenance manager without the required engineering review The licensee accepted the violation and responded in Letter TXX-93315 dated September 2, 1993. The pump was reworked thereby establishing.the required DCA configuration. The maintenance department issued a lessons learned on equipment design change requirements on October 8, 1993, and distributed it to all disciplines for review. This violation was considered to be an isolated instance and the licensee has taken appropriate corrective action t 7 FOLLOWUP (92701)
-(Closed) Inspection Followup Item 445/9214-03: Main Control Board Annunciators This inspection followup item identified a concern with the number of lit or inoperable annunciators which could desensitize licensed operators to changing plant condition The inspectors have been following the licensee's attempt to reduce the number of problem annunciators from a daily average of-20; however, problem 1 annunciators on both units still average approximately 20-25 on a daily basis :
while at 100 percent power. System engineering has eliminated the " mode"-
dependent alarms and this has greatly reduced the safety-related. annunciator problems. Continual daily emphasis has been placed on problem annunciator status at the plan-of-the-day meetings and by the maintenance organizations-to -
quickly troubleshoot and repair the annunciator Recently, TU Electric's management initiated an annunciator task team consisting of personnel from engineering, operations, I&C and electrical !
maintenance, work control, and planning and scheduling. New aggressive
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expectations for time out-of-service and repairs have been establishe Engineering has requested authorization to develop specifications for a new annunciator system, with a projected installation date of 1996. The goal for -
the new annunciator system is to reduce the number of lit annunciators to less than 20 for all modes and less than 5 for 100 percent powe To date, the licensee's efforts have not been effective in reducing the total number of problem annunciators; however, they have dedicated numerous :
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has developed realistic goals, and the inspectors will monitor the licensee's !
progress in meeting those goal :
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ATTACHMENT 1 1 PERSONS CONTACTED Licensee Personnel 0. Bhatty, Licensing W. J. Cahill, Jr., Group Vice President, Nuclear Engineering and Operations R. D. Calder, Engineering T. A. Hope, Site Licensing Manager J. J. Kelley, Vice President, Nuclear Engineering and Support .
D. C. Kross, Operations J. J. LaMarca, Manager, Engineering Outage B. T. Lancaster, Manager, Plant Support F. W. Madden, Mechanical Engineering Manager D. M. McAfee, Nuclear Operations J. W. Muffett, Manager of Technical Support & Design Engineering C. L. Terry, Vice President, Nuclear Operations A. Simon, Maintenance R. A. Smith, System Engineering D. Snow, Licensing M. W. Sunseri, Engineering The personnel listed above attended the exit meeting. In addition to the personnel listed above, the inspectors contacted other personnel during this inspection perio EXIT HEETING An exit meeting was conducted on September 30, 199 During this meeting, the inspectors reviewed the scope and findings of the report. When asked by the inspector if there were any questions or comments regarding the findings, none were raised. The inspectors confirmed with the licensee the corrective actions regarding the noncited violation and were assured that the actions were as stated. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspector i i
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- ATTACHMENT 2
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-EMPLOYEE CONCERNS PROGRAMS ,
y PLAhT NAME: Comanche Peak Steam Electric Station
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LICENSEE- TU Electric
. DOCKETS: 50-445; 50-446 ,
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A.- PROGRAM Does the licensee, have an employee concerns program? -*
(Yes or No/ Comments)
g YES !
~' Has NRC inspected the program?
YES, Reports 50-445/92-60; 50-446/92-60 and 50-445/88-23;.
50-446/88-20 1 SCOPE Is it for: ) Technical? (Yes, No/ Comments) .YES .!
.. i b.- Administrative? (Yes, No/ Comments) YES ;j I Personnel issues? (Yes,-No/ Comments) YES Does it cover safety-as well as non-safety issues?
(Yes or No/ Comments) Y.ES Is it designed for:
' Nuclear safety?' (Yes, No/Coments) --YES
- Personal safety? (Yes, No/Coments) YES- Personnel issues - including union grievances?
(Yes or No/Coments)
YES, BUT THE PROGRAM HAS NO AUTHORITY TO
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FILE GRIEVANCE ON BEHALF 0F THE' PERSON- Does the program apply to all licensee employees? l j
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(Yes or No/ Comments) YES Contractors?
(Yes or No/ Comments) YES Does the licensee require its contractors and their subs to have a similar program? (Yes or No/ Comments)
N EVERYONE BADGED ONSITE IS PROVIDED SAFETEAM ORIENTATION OR GIVEN INFORMATION ABOUT THE PROGRAM Does the licensee conduct an exit interview upon terminating .
employees asking if they have any safety concerns? (Yes or No/ Comments)
YES i INDEPENDENCE What is the title of the person in charge? MANAGER, SAFETEAM Who do they report to? MANAGER OF REGULATORY AFFAIRS Are they independent of line management? YES Does the ECP use third party consultants?
YE INVESTIGATORS ARE THIRD PARTY CONSULTANTS, '
SUBCONTRACTED TO UTILITY TECHNICAL SERVICES (UTS), THE OWNER OF THE PROGRAM How is a concern about a manager or vice president followed up?
THE ISSUE IS TAKEN TO AN INDIVIDUAL AT LEAST ONE STEP IN THE CHAIN OF COMMAND AB0VE THE PERSON IN QUESTIO THAT INDIVIDUAL HAS THE RESPONSIBILITY AND AUTHORITY TO ADDRESS AND RESPOND TO THE MATTER. AN INTERNAL REVIEW ORGANIZATION WHICH REPORTS TO THE CHAIRMAN, TU ELECTRIC, COULD ALSO BE CALLED IN TO FOLLOW UP ON SUCH A MATTE RESOURCES What is the size of the staff devoted to this program?
' PEOPLE (3 INVESTIGATORS, 5 INTERVIEWERS [1 FULL TIME, 4 PART TIME], PROGRAM MANAGER, AND SECRETARY) What are ECP staff qualifications (technical training, interviewing training, investigator training, other)?
MANAGER RECEIVED TRAINING FROM UTS ON OVERALL OPERATION OF THE PROGRAM. ALL INlERVIEWERS HAVE RECEIVED INTERVIEWING TRAININo INVESTIGATORS ARE TRAINED AND SUBCONTRACTED FROM UTS. TUE CORPORATE SECURITY INVESTIGATORS ARE CERTIFIED BY THE STATE-OF TEXAS. JOB DESCRIPTIONS AND CRITERIA ARE IN PLACE FOR ONGOING EVALUATION OF STAFF PERFORMANC '~ -3-
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-, REFERRALS Who has followup on concerns (ECP staff, line management, other)?
LINE MANAGEMENT, EXCEPT FOR THOSE ITEMS RELATED TO-NUCLEAR SAFETY, WHICH ARE HANDLED BY SAFETEAM INVESTIGATORS. CORRECTIVE ACTIONS ARE TRACKED THROUGH THE EXISTING PROJECT TRACKING SYSTEM. SAFETEAM PROGRAM 0FFICE PROVIDES A WRITTEN RESPONSE TO EVERY CONCER . CONFIDENTIALITY Are the reports confidential?
(Yes or No/ Comments)
YES Who is the identity of the alleger made known to (senior '
management, ECP staff, line management, other)?
(circle. if other explain)
ONLY THE ECP INTERVIEWER STAFF KNOWS THE NAME UNLESS THERE IS A NEED TO KNOW AND THE CONCERNED EMPLOYEE HAS SIGNED A RELEAS . Can employees be: , Anonymous? (Yes, No/Coments)
YES Report by phone? (Yes, No/ Comments)
YES
i FEEDBACK Is feedback given to the alleger upon completion of the followup? I (Yes or No - If so, how?) ;
YES, RESPONSE IS BY LETTE . Does program reward good ideas?
NO, ALTHOUGH APPRECIATION IS EXPRESSED TO PEOPLE FOR l
THE ISSUES RAISED, THESE FINDINGS ARE NOT PUBLICLY SHARED. REWARDING GOOD IDEAS IS NOT ONE OF THE ESTABLISHED OBJECTIVES OF THIS SPECIFIC PROGRA l Who, or at what level, makes the final decision of resolution?
THE SAFETEAM REVIEW COMMITTEE, CHAIRED BY THE MANAGER, SAFETEAM, WHO SIGNS ALL 0F THE RESPONSE LETTER . Are the resolutions of anonymous concerns disseminated?
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NO, BUT THE RESOLUTION MAY BE REVIEWED BY CONCERN NUMBER IF REQUESTE ) Are resolutions of valid concerns publicized (newsletter, bulletin ;
board, all hands meeting, other)? -l u
YES, THROUGH GROUP MEETINGS AS PART OF CORRECTIVE ACTIONS. THEY ARE NOT PUBLICIZED VIA NEWSLETTER OR BULLETIN BOARD H. EFFECTIVENESS - How does the licensee measure the effectiveness of the program?
UTS PERFORMS AN ANNUAL AUDIT OF THE PROGRAM. THE l QUALITY ASSURANCE ORGANIZATION PERFORMS AUDITS OF THE l PROGRAM. UNSOLICITED COMMENTS REGARDING THE PROGRAM- l l
ARE REVIEWED. A LOW PERCENTAGE (< 1/2 0F 1%) 0F CONCERNEES HAVE RECONTACTED THE OFFICE EXPRESSING DISSATISFACTION WITH THE RESOLUTION OF THEIR CONCER A REVIEW COMMITTEE OF DIVERSE MANAGERS, AS WELL AS SENIOR UTILITY MANAGEMENT, ALSO PROVIDES OVERSIGHT OF THE PROGRA . Are concerns: Trended? (Yes or No/ Comments)
YES, AND CATEGORIZE Used? (Yes or No/ Comments)
YES In the last three years how many concerns were raised? ]
1207 I
Of the concerns raised, how many were closed?
1195 What percentage were substantiated?
WHETHER A CONCERN IS SUBSTANTIATED IS ONLY TRACKED FOR l WHAT ARE TERMED " CLASS 1" CONCERNS. A CLASS 1. CONCERN )
IS ANY CONCERN INVOLVING NUCLEAR SAFETY, PLANT- !
OPERATION, ENVIRONMENTAL MATTERS, OR HARASSMENT / INTIMIDATION ISSUES. OF THESE ISSUES, 8.7% 1 (26/299) WERE SUBSTANTIATED OVER THE LAST THREE YEAR )
i How are followup techniques used to measure effectiveness (random i survey, interviews, other)?
THE CLOSING PARAGRAPH OF THE CLOSURE LETTER ASKS THE '
.CONCERNEE TO CONTACT THE SAFETEAM ORGANIZATION IF ANY-QUESTIONS REMAI . .
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I How frequently are internal audits of the ECP conducted and by whom?
INTERNAL AUDITS ARE PERFORMED ON REQUFET. UTS, THE l OWNER OF THE PROGRAM, PERFORMS AN ANNLAL AUDI ADMINISTRATION / TRAINING: Is ECP prescribed by a procedure? (Yes or No/ Comments)
YES .
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program (training, newsletter, bulletin board, other).?
POSTED SIGNS, FLYER HANDOUTS, GENERAL EMPLOYEE i
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TRAINING, PERSONNEL PROCESSING ORIENTATION, COMPANY ORIENTATIO ADDITIONAL COMMENTS: (Including characteris',ics which make the program especially effective, if any.) SENIOR MANAGEMENT'S APPARENT COMMITMENT TO THE PROGRA . CONFIDENTIALITY STRONG REVIEW COMMITTEE
, PROFESSIONAL HANDLING 0F CONCERNS l
' SEVERAL COMPANIES HAVE VISITED THE UTILITY TO LEARN ABOUT THE PR'SRAM AND TU ELECTRIC'S IMPLEMENTATIO NAME: L. A. YANDELL TITLE: CHIEF, REACTOR PROJECT SECTION B PHONE: (817) 860-8182 DATE COMPLETED: 8/19/93
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