IR 05000413/1987005
| ML20207T401 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/06/1987 |
| From: | Lesser M, Peebles T, Skinner P, Van Doorn P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207T389 | List: |
| References | |
| 50-413-87-05, 50-413-87-5, 50-414-87-05, 50-414-87-5, NUDOCS 8703230531 | |
| Download: ML20207T401 (10) | |
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Report Nos.:
50-413/87-05 and 50-414/87-05,i e4
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Licensee: Duke Power Company
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422 South Church Street
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Charlotte, NC 28242 Docket Nos.: '50-413 and 50-414 License"Nos.: NPF-35 and NPF-52
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Facility Name: Catawba 1 and 2
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Inspection Conducted: January 26 - February 25, 1987
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Inspectors:
I 4b MS/r,7 w
P. K. Van Dorrn Ofate Signed
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h P. }l. Skinner ~
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3/g/rp M.
S." Lesser
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Dat'e Signed Approved by:
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2 T. A.'Pe'ebles, Section Chief Date Signed
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Projects Branch 2 Division of Reactor Projects SUMMARY Scope: This routine, unannounced inspection was conducted on site inspecting in the areas of review of plant operations; surveillance observation; maintenance observation; review of licensee nonroutine event reports; review of IE Bulletins; and followup of previously identified items.
Results:
Of the six (6) areas inspected, two (2) apparent violations were identified in two areas (Failure to follow procedure resulting in violating Technical Specifications (TS) 4.0.2 and 4.0.4, paragraph 5.d; and Failure to meet the requirements of TS 3.6.3 for valve 2BB-8A, paragraph 6.c).
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REPORT DETAILS 1.
Persons Contacted Licensee Employees J. W. Hampton, Station Manager
- M. E. Anderson, Performance Associate Engineer
- H. B. Barron, Operations Superintendent
- W. F. Beaver, Performance Engineer
- M. Bolch, Emergency Planner W. H. Bradley, QA Surveillance
- B. F. Caldwell, Station Services Superintendent R. N. Casler, Operating Engineer R. H. Charest, Station Chemistry Supervisor
- M. A. Cote, Licensing Specialist
- T. E. Crawford, Integrated Scheduling Superintendent
- D. K. Davies, Projects Assistant Engineer W. P. Deal, Health Physics Supervisor C. S. Gregory, I&E Support Engineer C. L. Hartzell, Compliance Engineer J. Knuti 0perating Engineer F. N. Mack, Project Services Engineer W. W. McCollough, Mechanical Maintenance Supervisor C. E. Muse, Operating Engineer F. P. Schiffley, II, Licensing Engineer G. T. Smith, Maintenance Superintendent J. Stackley, !&E Engineer
- D. Tower, Shift Operating Engineer
- R. F. Wardell, Technical Services Superintendent
- R. White, Catawba Safety Review Group J. W. Willis, Senior QA Engineer, Operations Other licensee employees contacted included technicians, operators, mechanics, security force members, and office personnel.
- Attended exit interview 2.
Exit Interview e
The inspection scope and findings were summarized on February 25, 1987, with those persons indicated in paragraph 1 above.
The inspector described the areas inspected and discussed in detail the inspection findings.
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.
The following new items were identified at the exit interview:
Violation 50-414/87-05-03:
Failure to follow procedure resulting in
a violation of Technical Specification 4.0.2 and 4.0.4.
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Unresolved item 50-414/87-05-04:
Channel check on off scale high
Auxiliary Feed flow gauges.
Unresolved Item 50-413,414/87 05-01:
Management review and
corrective action of excessive problems occurring on a specific assigned shift.
Unresolved Item 50-413,414/87-05-02:
Potentially inadequately sized
value actuators.
Violation 50-414/87-05-05:
Failure to comply with Technical
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Specification 3.6.3 associated with inoperable containment isolation valve 2BB-8A.
3.
LicenseeActiononPreviousEnforcementMatters(92702)
a.
(CLOSED) Violation 413/86-43-01:
Failure to Maintain an Adequate Procedure for Conducting the Residual Heat Removal Pump 1B Performance Test While in Mode 6.
The response to this item was addressed by the licensee in correspondence dated December 5,1986.
The inspector reviewed the corrective action taken and considers this item as closed, b.
(CLOSED) Violation 413/86-30-02:
Failure to Follow Procedures While Attempting to Replace the 1A Seal Injection Filter.
The licensee addressed corrective action for this item in correspondence dated October 16, 1986. The inspector reviewed the corrective action taken and considers this item closed.
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Unresolved Items *
Three new unresolved items are identified in paragraphs 5.e., 6.b and c.
5.
Plant Operations Review (Units 1 & 2) (71707 and 71710)
a.
The inspectors reviewed plant operations throughout the reporting period to verify conformance with regulatory requirements. Technical Specifications (TS), and administrative controls. Control room logs, danger tag logs, Technical Specification Action Item Log, and the removal and restorction log were routinely reviewed. Shift turnovers were observed to verify that they were conducted in accordance with approved procedures.
The inspectors verified by observation and interviews, the measures taken to assure physical protection of the facility met current requirements.
Areas inspected included the security organization, the establishment and maintenances of gates, doors, and isolation zones in the proper condition, that access control and badging were proper and procedures followed.
- An Unresolved Item is a matter about which more information is required to determine whether it is acceptable or may involre a violation or deviation.
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In addition to the areas discussed above, the areas toured were observed for fire prevention and protection activities.
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included such things as combustible material control, fire protection systems and materials, and fire protection associated with mainte-nance activities.
The inspectors reviewed Problem investigation Reports to determine if the licensee was appropriately documenting problems and implementing appropriate corrective actions.
On February 19, the Resident Inspectors observed the licensee's
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annual Emergency Drill.
This is detailed in IE Report 50-413,414/87-07.
On January 28 Unit #2 experienced a reactor trip from 100% due to a 20 low-low steam generator level caused by a wire being pinched by a control cabinet door.
This wire caused a short in the feed regulating valve causing the valve to shut.
The plant was returned to power on January 30th and during the process 'of loading the turbine, at 21% power the turbine was tripped due to a high level in the moisture separator reheater.
The reactor war then shutdown and subsequently returned to power on February 1.
On January 30, 1987, Unit I was shutdown from 100% power due to loss of operability of both Containment Air Return and Hydrogen Skimmer Systems (See report 50-413,414/87-06).
Unit I was returned to power on January 31 at which time during loading of the turbine the unit tripped at 25%
power caused by a Power Range High Flux Low Setpoint being reached.
The plant returned to power operation on February 1.
b.
The inspector reviewed licensee control of overtime to assure that Technical Specification 6.2.2 administrative requirements were being met.
The inspector discussed procedural control, the approval process for overtime, and the amount of overtime being used with licensee management personnel in various plant groups including operations, instrument and electrical maintenance, mechanical maintenance, health physics, chemistry and performance.
In addition, the inspector generally reviewed the 1986 overtime reports for operations personnel and specifically calculated weekly overtime use for seven randomly selected individuals.
c.
On Jar.uary 29, 1987, NRC management (Messrs. Luis Reyes, Director.
Division of Reactor Projects, RII; Thomas Peebles, Project Section Chief, RII and Kahtan Jabbour, Licensing Project Manager, NRR) met with licensee management to tour the facility, review operational history, discuss present plant status and discuss performance indicators.
d.
On February 2, the licensee identified that during the shutdown of Unit 2 which occurred on January 30, 1987, to Mode 3 conditions, numerous semi-daily Technical Specification (TS) surveillances had not been performed.
Further investigation of this event identified
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f that OP/2/A/6100/02, Controlling Procedure for Unit Shutdown,
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Enclosure 4.1 step 2.14 requires that prior to entering Mode 3 PT/2/A/4600/19C, Pre-Mode 3 Surveillance Items had been completed.
PT/2/A/4600/19C step 12.2, requires a verification that a Mode 2 Periodic Surveillance Procedure (PT/2/A/4600/02B) had been completed within the previous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when entering Mode 3 from Mode 2.
An apparent comunication failure resulted in step 2.14 of OP/2/A/6100/02 being signed off as having been performed, when it had not been accomplished.
The failure to perform this step resulted in a failure to perform various semi-daily surveillance requirements. A review of the surveillance testing performed immediately preceding and following the missed surveillance was conducted and no changes to system and instrument operability were identified.
This failure to follow procedure resulted in a violation of TS 4.0.2 which requires each surveillance requirement to be performed within the specified time interval and TS 4.0.4 which requires that entry into an operational mode or other specified condition shall not be made unless the surveillance requirements associated with the limiting conditions for operations have been performed within the stated surveillance interval or as otherwise specified. This is identified as a violation 414/87-05-05; Failure to Follow Procedures Resulting in a Violation of TS 4.0.2 and TS 4.0.4.
e.
On January 31, 1987, Unit 1 experienced a power range high flux low setpoint reactor trip during power ascension shortly after paralleling the turbine generator to the grid.
Investigation into this event identified that the unit was at approximately 10% power when the operator closed the breaker and commenced loading the turbine at the rate of 10%/ minute with the load limit set at 100%
load.
As the load was being picked up on the turbine, a resulting decrease in primary temperature occurred.
The control room operator started withdrawing control rods in an effort to raise temperature.
Power reached the trip point of 25%, prior to the operator blocking this function, causing the reactor trip.
This excursion which occurred in approximately 3 minutes, appears to have resulted from an excessive rate of power change with an apparent lack of control of the evolution.
Present during the excursion were the turbine operator, the senior reactor operator and the shift supervisor.
On February 24,1987, Unit 2 tripped from 100% power. A preliminary investigation identified that the control room operator (CRO) noted on the operator aid computer (OAC) that bus 2EPD (125VOC vital instrumentationbus)hadanindicatedundervoltageconditionexisting on it.
This was discussed with shift supervision personnel.
A nuclear equipment operator (NED) was sent to physically check the position of the supply breaker for this bus. Upon investigation the NE0 reported to the CR0 that the breaker was open.
Supervision was reviewing to determine loads off this bus and when told the breaker was open, stopped the review process and directed the NE0 to reset and close the breaker.
Sirce the breaker was in fact not open, upon opening the breaker, the reactor tripped, since this bus supplied the
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undervoltage circuit of the train B reactor trip breakers. Recovery of this event was performed in accordance with applicable procedures.
The inspectors are concerned that both of these reactor trips must be added to a list of numerous significant events that have occurred during this one specific shift's activities.
This observation was identified to the inspectors by licensee management.
Further review of the problems associated with this particular shift will be performed by the licensee and NRC. This area is being identified as an Unresolved Item 413,414/67-05-01:
Management Review and Corrective Action of Excessive Problems Occurring on a Specific Assigned Shift, pending NRC review of licensee investigation and corrective actions.
One violation was identified as discussed in paragraph 5.d above.
6.
SurveillanceObservation(Units 1 & 2) (61726)
a.
During the inspection period, the inspector verified plant operations were in compliance with various TS recuirements.
Typical of these requirements were confirmation of compliance with the TS for reactor coolant chemistry, refueling water tank, emergency power systems, safety injection, emergency safeguards systems, control room ventilation, and direct current electrical power sources.
The inspector verified that surveillance testing was performed in accordance with the 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 requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The folliwing surveillance test were reviewed:
PT/1(2)/A/4200/02B Cold Shutdown Inside Containment Verification PT/1(2)/A/4200/02A Monthly Outside Containment Integrity Verification IP/0/A/3710/15 Quarterly Inspection on Battery 2EBA, 2 EBB The following surveillance test was observed:
IP/2/A/3222/00C Channel III Analog Operational Test b.
Technical Specifications (TS) 4.3.3.5 and 4.3.3.6 require a monthly channel check on Auxiliary Feed Flow Rate (CA) during Modes 1, 2 and 3.
ProblemInvestigationReport(PIR)2-C87-0031 identified that CA flow gauges on Uni'c 2 only are pegged high while at 100% power.
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This is because Unit 2 has 0-5 Steam Generators and approximately 13%
of total main feed flow is diverted through the auxiliary feed flow nozzles.
This large amount of flow peggs the 0-600gpm CA flow gauges.
The channel check is accomplished while performing PT/2/A/4600/03A, Monthly Surveillance Items.
On two occasions (8/26/86 and 1/4/87) the channel check was accomplished and determined acceptable by the licensee with the gauges pegged high off scale. This does not appear to meet the intent of a channel check as defined in Technical Specifications in that operability cannot be determined on an offscale instrument.
This also does not meet the acceptance criteria of the surveillance tests in that the channels are verified within 30gpm of each other.
The licensee has argued that a channel check is acceptable under these conditions because a pegged high gauge is expected at 100% power and if pegged high gauges are observed, the channel check is satisfactory. A larger scale flow transmitter and gauge would sacrifice required accuracy for auxiliary feed flow when needed.
The PIR has been forwarded to Duke design to verify this interpretation.
This is being identified as Unresolved Item 414/87-05-04: Channel Check on Off Scale High Auxiliary Feed Flow Gauges, pending determination of validity of the channel check, c.
On the evening of February 3,1987, Unit 2 was at or about 60%
reactor power conducting In Service Testing (IST) on valve 2BB-8A (Steam Generator 20 Blowdown Containment Inside Isolation Valve).
After several unsuccessful attempts to obtain a satisfactory isolation time (less than or equal to 10 seconds) the valve was listed as inoperable in the Technical Specification Action Item Log (TSAIL) at approximately 12:35 a.m.,
on February 4.
Technical Specification (TS) 3.6.3 requires the valve to be restored to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or isolation of the affected containment penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or for the reactor to be placed in 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 />.
The Catawba Nuclear Technical Specification Interpretation Manual provides clarification and additional guidance for operators on various Tech. Specs.
An interpretation for TS 3.6.3 dated June 11, 1985, provided incorrect guidance, allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for valve repair after isolation time criteria failure.
The TS interpretation was based on a May 8, 1985 letter to T. L. McConnell, Manager McGuire Nuclear Station from Duke Power Company Licensing.
The letter incorrectly applied provisions in IWP-3417 to permit a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> grace period to effect repairs to TS containment isolation valves which do not meet maximum isolation time acceptance criteria.
(The grace period was not allowed for valves which would not move to the intended position).
The operators thus assumed they had 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to comply with the action statement of TS 3.6.3.
Further questioning by the morning Shift Supervisor of licensing personnel revealed this particular TS interpretation to be invalid and 2BB-8A was closed at about 8:39 a.m., February 4.
This is identified as Violation 414/87-05-05:
Failure to Comply with Technical Specification 3.6.3 Associated with Inoperable Containment Isolation Valve 2BB-8 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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Historically the BB (Blowdown Containment Isolation Valves) on both units have had isolation times in the 9 to 10 second range, the maximum allowable time being less than or equal to 10 seconds. The problem apparently is the valve actuators have been designed to close in 10 seconds thus leaving little tolerance margin.
Apparently on more than one occasion, a valve has failed the 10 second criteria upon the first stroke. This may not have been indicative of actuator degradation but simply a tolerance problem between design values and acceptance criteria.
This problem may additionally exist with other containment isolation valves in that valves are routinely unable to meet acceptance criteria.
This is being identified as Unresolved Item 413,414/87-05-02:
Potentially inadequately Sized Valve Actuators, pending licensee review of the extent of affected valves and resolution.
One violation was identified as described in paragraph 6.c above.
7.
Maintenance Observations (Units la 2)(62703)
a.
Station maintenance activities of selected systems and components were observed / reviewed to ascertain that they were conducted in accordance with requirements.
The inspector verified licensee conformance to the requirements in the following areas of inspection:
the 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 detemine status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may effect system performance.
The following maintenance activities were observed:
Various portions of diesel generator 2A lube oil and fuel oil filter replacement.
The following maintenance work requests were reviewed:
SWR 5406 Diesel Engine Fuel Oil Filter Remov,1 and Replacement MNT 4196 Diesel Engine Head Gasket Leak SWR 5408 Diesel Engine Fuel Oil Strainer MNT 4199 Crankcase Relief Valve Oil Leak No violations or deviations were identifie %
8.
Review of Licensee Nonroutine Event Reports (Units 1&2)(92700)
The below listed Licensee Event Reports (LER) 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 potential generic problems, reporting requirements satisfied, and the relative safety significance of each event. Additional inplant reviews and discussion with plant personnel, as appropriate, were conducted for those reports indicated by an (*).
The following LERS are closed:
- LER 413/86-44, Rv. 1 Both Trains of Residual Heat Removal Inoperable Due to Personnel Error
- LER 413/86-57 Rv. 2 Inadequate Valve Operator Torque Settings Due to Manufacturer Deficiency
- LER 413/86-59, Rv. 1 Containment Pressure Channel Unknowingly Inoperable Due to Personnel Error LER 414/86-50 Termination of Containment Air Release Due to Conservative Radiation Monitor Setpoint LER 414/86-52 Auxiliary Feedwater Auto-Start Signal on Loss of Main Feedwater Pumps on Windmill Protection Due to Unknown Cause No violations or deviations were identified.
9.
IEB 85-01 - Steam Binding of Auxiliary Feedwater Pumps (Units 1 & 2)
(92703)
The licensee responded to this bulletin by correspondence dated February 25, 1986.
Procedures were developed to monitor fluid conditions in the auxiliary feedwater piping)/A/6250/02)once per day using a hand held pyrometer.
This procedure (0P/1(2 also included information to recognize steam binding and to restore the system to operable status.
Hardware modifications have been made to the auxiliary feedwater system to install thermocouples that provide temperature instrumentation with alarm capability to identify leakage past the check valves. The procedures are developed and approved in accordance with the licensee's commitments and requirements.
Personnel have been trained in the use of these procedures and how steam binding is precluded.
Based on this review this item is closed.
No violations or deviations were identifie,, _.
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1 10. Previously Identified Inspector Findings (92701)
(CLOSED)InspectorFollowupItem 413/85-28-1: Retain ECP Calculations. A review of this item identified that procedures controlling surveillance requirements for unit startup requires a copy of the reactivity balance calculation to be retained.
Based on this review this item is closed.
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