IR 05000413/1992029
| ML20127E854 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/06/1993 |
| From: | Belisle G, Hopkins P, William Orders, John Zeiler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20127E815 | List: |
| References | |
| 50-413-92-29, 50-414-92-29, NUDOCS 9301200067 | |
| Download: ML20127E854 (16) | |
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Report Nos.:
50 413/92-29 and 50-414/92-29 Licensee:
Duke Power Company 422 South Church Street Charlotte, N.C.
28242 Docket Nos.:
50-413 and 50 414 License Nos.:
NPF-35 and NPf-52
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facility Name:
Catawba Nuclear Station Units 1 and 2 Inspection Conducted: Nove:aber 8,1992 December 12, 1992
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GA J.~ETler, Resi ent[ lits ector
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Inspector:
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Accompanying Personnel:
S. M. Congel e
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Projects Section 3A Division of Reactor Projects SUMMARY-Scope:
This routine resident inspection included, but was not limited to the-following areas; plant operations. review, Auxiliary feedwater (CA) System design problem review,. improper use of chemistry procedure review, review of containment temperature deviation, surveillance observation review, maintenance observation review, review of VOTES system inaccuracies, review of diesel generator cooling water heat. exchanger failures, review of Licensee Event Reports, and followup on previous inspection findings.
Results:
One violation was identified involving electrical circuitry design errors in'the Auxiliary feedwater Systems on both units which would have prevented the systems from performing their-
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-2 intended-safety function during-certain-design basis-events (paragraph 4.0).
Two Non-Cited Violations'were identified involving the failure to
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follow Chemistry Department procedures (paragraph 5.0) and a Mode change which was made on Unit 2 with containment temperature below the technical specification required limit (paragraph 6.0).
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REPORT DETAILS 1.
Persons Contacted Licensee Employees S. Bradshaw, Shift Operations Manager
- J. Cox, Acting Regulatory Compliance Manager
- J. Forbes, Engineering Manager S. Frye, Operations Support Manager E. Geddie, Operations Superintendent T. Harrall, Safety Assurance Manager
- J.
Lowery, Compliance
- W. McCollum, Station Manager K. Seasely, Compliance M. Tuckman, Catawba Site Vice-President Other licensee employees contacted included technicians, operators, mechanics, security force members, and office personnel.
NRC Resident Inspectors
- W. Orders P. Hopkins
- J Zeiler Accompanying NRC personnel
- S. Congel
- Attended exit interview.
2.
Plant Status Unit 1 Summary Unit 1 operated the entire report period at essentially full power.
Unit 2 Summary Unit 2 began the report period at full power.
On November 15, the unit was shut down to repair the seals on Reactor Coolant Pump 2D which had experienced an increase in No.1 seal leakoff.
Seal repairs were completed on November 22.
Reactor Startup commenced on November 29 and the unit was placed on-line the following day. The unit reached full power on December 3 and operated unencumbered for the remainder of the report period.
3.
Plant Operations Review (71707)
a.
The inspectors reviewed plant operations throughout the report i
period to verify conformance with regulatory requirements,
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Technical. Specifications (TS) and administrative controls.
Control Room logs, the Technical Specification Action Item Log, and the Removal and Restoration (R&R) log were routinely reviewed.
Shift turnovers were observed to verify that they were conducted
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Further, daily plant-status meetings were routinely attended.
Plant tours were performed on a routine basis.
The areas toured included but were not limited to the following:
Turbine Buildings Auxiliary Building Units 1 and 2 Diesel Generator Rooms Units 1 and 2 Vital Switchgear Rooms Units 1 and 2 Vital Battery Rooms Standby Shutdown Facility Activities prescribed in the following operations procedures were reviewed in detail:
OP/0/A/6100/06 Reactivity Balance Calculations OP/0/A/6350/13 Placing Source Range Detector in Service OP/2/A/6150/08 Rod Control 0P/2/B/6300/01 Turbine Generator Shutdown OP/2/B/6400/01A Condenser Circulating Water System OP/2/A/6450/17 Containment Air Release and Addition System During the plant tours, the inspectors verified by observation and interviews.that measures taken to assure physical protection of the facility met current requirements. Areas inspected included the security organization, the establishment and maintenance of gates, doors, and isolation zones in the. proper conditions, and that access control badging' were proper and procedures followed.
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In addition, the areas toured were observed for fire prevention and protection-activities and radiological control practices.
The inspectors also reviewed Problem Investigation Reports (PIRs) to determine if the licensee was appropriately documenting problems and implementing corrective actions.
b.
TS 3.0.3 Entry Due to Control-Room Ventilation Inoperability On December 3, both units were operating at full power. At 10:00 a.m. that morning,.both units were placed under the' action requirements of TS 3.0.3 when both_ trains of;the Control Room-Ventilation System (VC/YC) were declared inoperable. Train A of the system had been removed from service for routine cleaning of-'
the chilled water condenser tubes on November 30. On the morning.
of December 3, an alarm was received on Train B indicating high w
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chili water temperature. Ultimately, a problem with the actuator motor for the chiller control vanes was identified which resulted in Train B being declared inoperable.
TS 3.0.3 was exited at 2:35 p.m., following the restoration and successful testing of Train A.
On December 6, Train B of the VC/YC System was returned to service following the repair of the actuator motor for the chiller control vanes and repair of the chill water pump bearings which were also found to be damaged.
No violations or deviations were identified.
4.
Auxiliary Feedwater (CA) System Design Deficiencies (71707)
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Suction Source Realignment Event Description:
On December 1, 1992, an instructor in the licensee's operator training department requested assistance from the on-site Component Engineering group (CES) in determining the effects of a hypothetical malfunction of a non-safety related DC power supply.
In a review of applicable electrical drawings, an engineer determined that the power supply in question powered portions of the electrical logic, which during selected accident scenarios, reali;ns the suction of both units' CA system turbine driven pumps (TDAFWPs) from the non-safety related, to the safety related suction source supplied by the Nuclear Service Water (RN) System.
By 7:30 p.m. that evening, the licensee had concluded that the motor driven CA pumps (MDAFWP) were not affected by this design flaw, but that the TDAFWPs on both units were inoperable in that they were incapable of automatically realigning to the safety related suction source. This placed Unit 1 in a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> action statement per TS 3.7.1.2 since the 1A motor driven CA pump was already out of service. Unit 2 on the other hand was placed in a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action statement for one it. operable CA pump.
Subsequently, the licensee's engineering staff concluded that by installing an electrical jumper around a set of contacts in the safety related circuitry controlling the aforementioned suction realignment, they could return the TDAFWPs to operable status.
The jumpers were installed as temporary station modifications (TSM) on both units. Unit I was completed by 1:15 a.m. the following morning and by 4: 16 a.m. on Unit 2, returning both TDAFWPs to operable service.
Background:
The CA System is designed to provide the safety function of supplying a safety related source of emergency feedwater to the steam generators (S/Gs) to maintain a secondary side heat sink at times when the normal feedwater system is not available.
Technical Specification Section 3.7.1.2 requires both trains of Motor Driven Auxiliary Feedwater Pumps (MDAFWP) and the Turbine
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Driven Auxiliary feedwater Pump (TDAFWP) to be operable in modes 1, 2, and 3.
The CA system is required to function for the following Design Basis Events:
Loss of Normal Feedwater flow, Loss of Offsite Power, Main Feedline Rupture, Main Steamline Rupture and Loss of Coolant Accident.
The design basis of the CA system requires the system to be able to withstand an initiating event, a single active failure and still meet the minimum secondary side heat sink requirements for the events listed above.
As a minimum, CA flow to two intact S/Gs is required to provide the minimum heat sink.
No credit can be taken for the use of non-safety equipment in mitigating the accident.
The normal suction alignment for the CA System is to the non-safety related condensate grade sources comprised of the CA condensate storage tank (CACST), the upper surge tank (UST) and the condenser hotwell. During a Design Basis Event, the potential exists to lose the non-safety sources after which a realignment to the safety related, assured suction source (RN) would take place using circuitry designed to detect decreasing suction pressure.
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The three permissives required to initiate the realignment to RN are an indication that the TDAFWP is running, low suction pressure in the CA suction header, and the presence of a TDAFWP auto-start signal.
Discussion:
A review of applicable electrical drawings limited the problem to i
the TDAFWP running permissive for the realignment logic.
This-signal is developed in a non-safety circuit.
This circuit has two
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non-safety inputs, one from each TDAFWP steam admission valve and one from the TDAFWP trip & throttle valve. According to the licensee's analysis'of the design basis events for which the CA
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system is required to respond, only the main feedline rupture
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accident appears to be adversely affected.
For this analysis, the i
A S/G main feedline is assumed to break between the S/G and the discharge check valve and the B MDAFWP is assumed to be the active i
single failure, in this scenario, all the flow from the running MDAFWP would feed the faulted A S/G due to the pump runout circuitry which would isolate flow to the B S/G.
CA system flow is required to two intact S/Gs to insure an adequate secondary side heat sink.
With the assumed failure of the non-safety l
components, the condensate grade sources are not available and the
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failure of the aforementioned non-safety circuit prevents the r
TDAFWP from realigning to RN. The CA system would not have been able to meet its design basis flow requirements for this accident, and thus would have been incapable of performing its safety function.
Safety Significance:
As part of their evaluation _ to determine the safety significance of this event, the licensee evaluated the probability of the above l
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described set of circumstances which would render the CA system incapable of performing its required function. The licensee calculated probability of these events occurring coincidentally to be on the order of IE-7/yr (excluding the failure of the condensate sources and the non-safety circuit).
Failure of the condensate grade sources must also be postulated which removes the normal suction source for all three CA pumps. According to the licensee, a seismic event is the only credible event which would fail the condensate sources.
Conventional analysis does not assume a seismic event coincident with a feedline break, however, for this accident scenario the licensee is performing an analysis which will evaluate the above failures along with a seismic event and loss of offsite power (LOOP).
Although the results of this
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analysis were not available at the end of this inspection, the licensee indicates that the probability of this sequence of events occurring coincidentally would decrease below IE-7/yr.
Conclusion:
The CA systems on both units have been inoperable since start-up in that they would not have been able to perform their intended safety function for the main feedline break Design Basis Accident.
However, the licensee's analysis indicates that the probability of this sequence of events occurring coincidentally is below lE-7/yr, and that the risk to the public was not significantly increased by the problem, b.
Speed Control Event Description On December 8, the licensee's engineering staff detected another related design deficiency involving the speed control circuitry on the TDAFWPs on both units.
The problem involves non-safety related circuitry located upstream of a solenoid valve which receives only an A train actuation signal and supplies control air to the TDAFWP governor. Upon receiving an auto start signal, the solenoid valve functions to isolate control air to the governor which in turn demands full turbine speed.
The licensee's engineering staff concluded that by opening an electrical link in the control power to the aforementioned solenoid valve, they could return the TDAFWPs to operable status.
The links were opened on both units using the aforementioned TSM process.
The TSM was implemented on both units by 2:00 A.M. on December 9.
Background:
For CA system background information refer to paragraph 4.a above.
Discussion:
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The licensee's analysis revealed that the CA system would not have been able to combat the design basis events of a feedwater or steam line break on S/G D.
The scenario in question involves an initiating event which includes a feedwater or steamline break on S/G 0, and a failure of A Train SSPS which is considered to be the active single failure.
This single failure would lead to the failure of A MDAFWP to start.
All flow from the B MDAFWP would be delivered to the faulted D S/G assuming the proper functioning of installed pump runout circuitry which would isolate CA flow to the C S/G.
This in turn means that TDAFWP would be relied upon to deliver the required flow.
The problem identified was the fact that the TDAFWP governor will
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not fail to the desired maximum speed setting on a B Train auto-start signal.
Failure of Train A SSPS (the active failure) would then result in the governor failing to its minimum speed setting assuming a malfunction of the non-safety related speed control circuitry.
At minimum speed, the TDAFWP will not develop sufficient driving head to deliver flow to the intact S/Gs. Thus a single non-safety failure could cause the TDAFWP governor to fail to the minimum speed and ultimately result in the CA system being incapable of performing its intended safety function.
Safety Significance:
The licensee performed an analysis to estimate the probability of the sequence of events occurring coincidentally. The analysis concluded that the frequency of core damage from sequences involving the failure of this non-safety circuit is less than 1.2E-8/yr.
Conclusion:
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The licensee's analysis concludes that the CA systems for both Catawba units had been inoperable since start-up in that were not capable of performing their intended safety functions to mitigate the consequences of a feedwater line or steam line break on the D S/G. The analysis also indicated however that the risk to the public was not significantly increased by the problem.
Regulatory Requirements:
The CA System is required to be designed in part such that the system can mitigate the consequences of a Design Basis Event, i.e., perform its intended safety function, assuming a single active failure.
Since no credit may be taken for non-safety equipment in mitigating a design basis accident, the CA systems for both units would not have been capable of performing their intended safety function for certain postulated Design Basis Accidents.
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10 CFR 50, Appendix B, Criterion 111, Design Control, requires in part that measures be established to assure that applicable regulatory requirements and the design basis specified in the license application for the facility be correctly translated into specifications, drawings, procedures, and instructions.
Further, design control measures are to provide for verifying the adequacy of design, by the performance of design reviews, the use of alternate or simplified calculational methods, or the performance of a suitable testing program.
Technical Specification 3.7.1.2 requires at least three Auxiliary Feedwater (CA) pumps and associated flow paths be operable in Modes 1-3.
Operability includes the capability of the system to perform its intended safety function during design basis events, assuming a single active failure, and with no credit taken for non-safety equipment toward mitigating the event.
Contrary to those requirements, design control measures associated with the CA System were inadequate in that design deficiencies have existed on both units since startup which rendered the systems inoperable in that they were not capable of performing their intended safety functions. The design deficiencies involve the interaction of non-safety related equipment with safety related control circuitry affecting the operation of both units'
Turbine Driven Auxiliary Feedwater Pumps (10AFWPs) such that during certain design basis events, malfunction of the non-safety related equipment, coupled with a single active failure would have prevented the CA Systems of bcth units from delivering the required flow to two intact S/Gs as designed. This issue is documented as Violation 413,414/92-29-01: Electrical Design Errors in the Auxiliary Feedwater System.
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One violation was identified.
5.
Failure to Follow Chemistry Procedures (71707)
On November 2, an inspector observed two chemistry technicians adding chemicals (corrosion inhibitors) to both Unit 2 Diesel Generator Engine Cooling Water Systems (KD).
The activity was to have been performed using procedure, OP/2/A/6400/25, Operating Procedure for the Addition of Chemical to Safety Related Closed Cooling Water Systems.
During the conduct of the activity, the individuals separated, with one individual performing procedural tasks on one diesel's KD system and the other individual performing similar tasks on the opposite diesel.
The inspector noted that there was only one copy of the procedure at the work location and that one of the individuals performed some portions of the activity, such as manipulation of several valves without the possession and in hand use of the procedure.
The inspectors discussed the above concern as well as other chemistry procedure issues including ger.eral lack of signoff steps, with Chemistry Department management.
In response, the valves which were manipulated
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by the two individuals were verified to have been returned to their proper alignment.
In addition, Problem Investigation Process (PIP) No.
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2-C92-0882 was initiated to investigate the praced"ral problems.
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i The in:pectors reviewed Chemical Management Frocedure (CMP) 2.2.'.4, Use of Procedures, which describes the requirements for use of chemistry department procedures.
Step 7.11 requires that procedures or enclosures be in the possession of a person performing the tasks delineated in a j
safety related procedure.
OP/2/A/6400/25, Operating Procedure for the Addition of Chemical to Safety Related Closed Cooling Water Systems is a
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safety related procedure.
Technical S)ecification 6.8.1 requires in part that written procedures be establis1ed, implemented and maintained covering the activities
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referenced in Appendix A of Regulatory Guide 1.33 Revision 2, February
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This failure to follow procedure CMP 2.2.14 is ;onsidered a
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violation of TS 6.8.1, however, af ter review of the circumstances relative to the issue, it was determined that the crneria specified in Section Vll.B.(1) of the NRC Enforcement policy were satisfied.
The violation was not willful, nor similar to a prior violation for which corrective actions have not been sufficient to prevent recurrence, and appropriate corrective action was initiated prior to the end of the o
report period.
For those reasons, this issue is documented as Non-Cited Violation (NCV) 414/92-29 02:
Failure to Follow Chemistry Procedure.
One NCV was identified.
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6.
Mode Cl with Containment Temperature Below Requirements (71707)
On Novemb.c 29, Catawba Unit 2 was in Mode 2, in the process of unit start-up following completion of a shutdown to repair a reactor coolant
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pump seal.
At 9:05 p.m. that evening,.the unit entered Mode 1.
At 9:30 p.m., a control room operator observed that the upper containment-average temperature was 71*F.
Technical Specification 3.6.1.5 requires that the upper containment average temperature be above 75'F when the unit is in Moda 1.
In response to the -identification of the problem, the control room operators isolated cooling water to the upper containment ventilation units, causing containment temperature to increase.
By 2:35 a.m. the following morning, temperature had returned to within TS limits for Mode 1 operation.
This event will be reported to the NRC in Licensee Event Report (LER) No. 414/92-05.
The licensee reviewed the Pre-Mode 1-surveillance procedure which was to have been performed-by the control room operators to verify compliance with TS curveillance items prior to the unit entering Mode 1.
It was discovered that the operator completing the procedure had-incorrectly signed-off that the upper containment temperature was greater than 78'F.
The actual temperature had been 73*F when the procedure was completed.
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It was also discovered that-the control room operators had failed to:
acknowledge a computer alarm received at 8 dh p.m. on November 29 indicating that upper containment temperature was outside_TS limits.
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Planned licensee corrective action inclu4s revising all the TS Pre Mode and Mode surveillance procedures to require that the astual value of the t
TS surveillance item being verified be entered instead of initialinu
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that the surveillance item is within acceptable limits.
In additica, an
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audible function may be added to the alarm on upper containment
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temperature, as well as other critical TS items.
The inspector hiso
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reviewed the licensee's evaluation for the effect of the low upper containment temperature on containment pressure following a LOCA. The evaluation confirmed that the resulting peak containment pressure would
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not have been significantly affected.
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This incident involved a violation of the requirements of TS 3.6.1.5
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and TS 3.0.4, for entry into Mode I without the conditior. -for the i
limiting Condition for Operation being met for that mude.
Atter review of the circumstances relative to-the issue however, it was determined that the criteria specified in Section Vll.B.(2) of the NRC Enforcement
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policy were satisfied.
The violation was licensee identified, ens not
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willful, will be reported, and, appropriate corrective action was initiated prior to-the end of the report period.
Accordingly this issue is documented as NCV 414/92-29 03:
Mode Change with Containment i
Temperature Below 15 Requiremen's.
One NCV was identified.
7.
Surveillance Observation (61726)
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a.
General During the inspection period, the inspectors verifled plant operations were in compliance with various TS requirements.
Typical of these requirements were confirmation of compliance with the TS for reactivity control systems, reactor conlant systems, safety injection systems, emergency safeguards systems, emergency power systems, containment, and other important plant support systems.
The inspectors verified that:
surveillance testing was
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performcd in accordance with approved written procedures, test instrumentation was calibrated, limiting conditions for operation were met, appropriate removal and restoration of the affected equipment _was accomplished, test results met acceptance criteria and were reviewed by personnel other than the individual directing the test, and any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel, b.
Surveillance Activities Reviewed
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The inspectors witnessed or reviewed-the following surveillances:
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PT/1/A/4200/18 Safety injection Power Disconnect Test PT/1/A/4200/05A Safety Injection Pump 1A Performance Test PT/i/A/4250/02B Weekly Main Turbine Valve Movement-
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l PT/1/A/4250/03B Auxiliary feedwater Motor Driven Pump 1B Performance Test PT/1/A/4250/06 Auxiliary feedwater Pump Head and Valve Verification PT/1/A/4350/02A Diesel Generator IA Operability Test
PT/1/A/4450/03A Annulus Ventilation System Train lA Operability Test PT/2/A/4150/01D Reactor Coolant System Leakage Calculation i
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PT/2/A/4150/02 Auxiliary Building filtered Exhaust System Operability PT/2/A/4200/41A Containment Purge Isolation Valve Leak Rate Test PT/2/A/4350/028 Diesel Generator 28 Operability Test PT/2/A/4600/17 Surveillance Requirements for Unit 2 Shutdown No violations or deviations were identified.
8.
MaintenanceObsenvations(62703)
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General Station maintenance activities of selected systems and components were observed / reviewed to ensure that they were conducted in accordance with the applicable requirements.
The inspectors verified licensee conformance to the requirements in the following areas of inspection:
activities were accomplished using approved procedures, and functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities performed were accomplished by qualified personnel; and materials used were.
properly certified. Work requests were reviewed to determine the status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which may affect system performance, b.
Maintenance Activit.es Reviewed The inspectors witnessed or reviewed the maintenance activities associated with the following Work Order. Tasks (W0s) or Instrumentation or Mechanical Procedures-(IPs or MPs);
WO 58919 OPS Repair-IKG-5, Will Not Control KG
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Temperature Properly WO 92084275-01 Perform Torque Testing on 1RN 849B WO 92085473-01 Inspect and-Repair DG2B Fuel Rack Assembly WO 92085717-01 Replace Spring Packs on RN Valve 849B WO 92086745 01 Replace Shutdown Cylinder on 28 DG WO 92090377-01 Investigate / Repair Valve 2SV-13 Stroke Time failure IP/0/A/3240/15 High Flux at Shutdown Alarm Adjustment.
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IP/0/A/3820/01 Limitorque Valve Operator Corrective Maintenance IP/0/M3820/03C Limitorque Actuator Testing Using Rotork Equipment IP/
20/0?G Limitorque Spring Pack Testing Using the Spring Pack Tester IP/0/A/3820/04 Operating Check Out of Limitorque and Rotork Valve Actuators IP/0/A/3890/08C Controlling Procedure for Wire Termination and Splicing (600 Volts or Less)
MP/0/A/7400/02 DG Governor Oil Cooler and Shutdown cylinder Corrective Maintenance MP/0/A/7650/02 Lubrication of Safety Related Equipment
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c.
VOTES System Inaccuracies
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On November 23, the inspectors received a copy of Catawba Problem investigation Report (PIR) 0 C92-0017 which deals with recently detected inaccuracles of the equipment and methodology of certain motor nperated valve test equipment supplied by Liberty Techno.ogies.
Attached to the PIR was a PART 21 notification issued by Liberty Technologies related to two issues concerning thrust values for motor operated valves (MOV) that are calibrated using that company's VOTES system.
The issues concern the possible use of improper stem material constants, and the failure to account for a torque ef fect when the calibrator is placed on the threaded portion of a small diameter, high-lead threaded valve stem.
Both issues cause the indicated thrust to be less than the true thrust.
On November 24, the inspectors discussed the issue with the appropriate parties in the licensee's engineering staff to determine if there were equipment operability concerns associated with the above mentioned inaccuracies.
It was concluded that there were no immediate concerns in that the inaccuracies ultimately result in more closing thrust being applied than assumed.
For those valves which have a safety function of closing, this is conservative, further, according to the engineering staff, these inaccuracies are not applicable to valves which have a safety function of opening.
The licensee is currently awaiting additional information from Liberty Technologies which is needed to perform a complete reanalysis of all MOVs which are tested using the VOTES system.
The inspectors will continue to follow this issue until the analyses and operability evaluations are complete.
This item will be tracked as Inspector followup Item (IFI) 413, 414/92-29-04:
Review Licensee Analysis of V0TES System Inaccuracie ~
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Diesel Generator Cooling Water Heat Exchanger failures On October 31, 1992, Catawba Unit 2 was operating at full power when at 8:17 p.m. that evening an alarm was received indicating a
low temperature in the cooling system of the 2A diesel generator
!
(D/0).
Subsequent investigation revealed that there was a tube
,
'
leak in the D/G cooling water (KD) heat exchanger, resulting in nuclear service water (R/N) leaking into the D/G engine jacket
,
water.
The D/G was declared inoperable that evening in order to effect repairs to the heat exchanger.
It was determined that one tube had experienced a circumferential shear as evidenced by visual examination. Ultimately, the tube was plugged and the D/G returned to service on November 2.
The inspectors noted that there was no root cause analysis performed tu determine why the tube had failed.
In discussions with the licensee, the inspectors were told that based on past non destructive examination of the heat exchanger, there was no reason to believe that the sheared tube was not a random, isolated event and that the heat exchangers on both D/G's would undergo thorough evaluations during the refueling outage scheduled to begin in early February,1993.
On November 17, Catawba Unit 2 was in mode 5 having been shut down the day before to effect repairs to the D reactor coolant pump seal s.
At approximately 6:00 p.m. that evening, it was determined that water from an unknown source was leaking into toe 2B D/G room
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sump.
Chemistry analysis revealed that the water was leaking from
the KD system. Ultimately, it_was determined that there was a tube leak in the KD heat exchanger.
Similar to the tube which had
,
failed in the 2A heat exchanger, the crack was transverse and had propagated 75% of the tube circumference.
A section of the tube was removed from the heat exchanger and sent to the 1_icensee's materials lab for analysis.
The tube location was plugged and the
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D/G was returned to service at 4:05 a.m. the following morning.
This issue will continue to be monitored and will be tracked as an IFI pending review of the laboratory analysis and heat exchanger inspections during the upcoming refueling outage.
This issue will be documented as IFl 414/92-29 05:- Review KD Heat Exchanger Tube Failure Analysis and Heat Exchanger Inspection During Upcoming Refueling Outage.
No violations or deviations were identified.
9.
Review of Licensee Event Reports (92700)
The below listed LERs were reviewed to determine if the information:
provided met NRC requirements.
The determination included: adequacy of description, verification of compliance with Technical Specifications and regulatory requirements, corrective action _taken, existence of-m
.
____. _ _ _. _. _ _ _
'.
,
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,
potential generic problems, reporting requirements satisfied, and the relative safety significance of each event, t
I a.
(Closed) LER 413/91-14: TS Violation Due to improper
,
Overtemperature Delta-Temperature Calculation.
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'
This issue involved an improper gain setting in the Overtemperature Delta-Temperature (OTDT) circuitry which prevented this reactor trip feature from functioning as intended over its
>
entire range.
The condition existed since startup on both units.
'
As immediate corrective action, the problem was corrected by changing the gain setting associated with the T-ave portion of the 010T circuitry.
Subsequent actions included revising the station procedure used to manually calculate the 010T trip set)oint.
The
'
inspector verified that these procedure revisions had acen performed.
The licensee's corrective actions were determined to be acceptable, b.
(Closed) LER 413/91-16: TS Violation as a Result of the Control Rod Position Alarm Monitor Being Inoperable Due to Management i
Deficiency.
This issue involved a logic error discovered in the Operator Aid Computer (OAC) Rod Position Alarm Monitor (RAM) program which monitors the Control Bank and Shutdown Bank rod positions and
provides alarms when rod positions are incorrect or if rods are moved in an unacceptable manner.
The logic error could have resulted in the failure to alarm conditions of misaligned rods.
As-immediate corrective action, the OAC RAM program was corrected.
Since there existed no formal policy requiring functional verification of the OAC programs during initial startup of the units,-all critical 0AC programs were reviewed by the licensee to ensure the programs operated as originally intended.
The
.
inspector considered the licenses's corrective actions to be i
adequate.
No violations or deviations were identified.
10.
Followup on Previous Inspection Findings (92701 and 92702)
L (Closed) Unresolved Item (URI) _ 413,414/92-2_7-02:
Review of Licensee Analysis of-Standby Makeup Pump Damper Pressure, 1.
in a-previous inspection,.the inspectors became aware of_ instances where-the Nitrogen gas charge in the pulsation dampers of both Unit's Standby Makeup Pumps (SMPs) had been found less than the vendor recommended pressure.
Based on past operability concerns by the inspectors, this issue was left unresolved pending the 1.icensee's analysis of_ the impact
-
of low pulsation damper pressure on the SMPs._
During this report period, the licensee completed the aforementioned past operability evaluation. The results of-this evaluation
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demonstrated the capability of the SMPs to perform their design function under the decreased damper pressure conditions identified.
The inspectors reviewed the evaluation and determined that the licensee had adequately addressed the past operability concerns.
At the end of the report period, the licensee was still pursuing long term resolution of the problem of losing damper pressure.
The inspectors will continue to monitor the licensee's effort to improve the reliability of the dampers.
No violations or deviations were identified, 11.
Exit Interview The inspection scope and findings were summarized on December 15 and 17, 1992, with those persons indicated in paragraph 1.
The inspector described the areas inspected and discussed in detail the inspection findings listed below. No dissenting comments were received from the licensee.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
Description and Reference i
Jtem Numbel VIO 413, 414/92-29-01 Electrical Design Errors in the Auxiliary Feedwater System (paragraph 4).
NCV 414/92-29-02 Failure to follow Chemistry Procedure Use Guidance (paragraph 5).
NCV 414/92-29-03 Mode Change with Containment Temperature Below TS Requirements (paragraph 6).
IFl 413, 414/92-29-04 Review Licensee Analysis of V0TES System inaccuracies (paragraph 8.c).
IFI 414/92-29 05 Review KD Heat Exchanger Tube failure Analysis and Heat Exchanger Inspection During Upcoming Refueling Outage (paragraph 8.d).
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