IR 05000254/1989016
| ML20246K998 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 08/28/1989 |
| From: | Lerch R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20246K991 | List: |
| References | |
| 50-254-89-16, 50-265-89-16, NUDOCS 8909060200 | |
| Download: ML20246K998 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-254/89016(DRP); 50-265/89016(DRP)
Docket lios. 50-254; 50-265 Licenses No. DPR-29; DPR-30 Licensee: Commonwealth Edison Company i
Post Office Box 767 I
Chicago, IL 60690 Facility Name: Quad Cities Nuclear Power Station, Units 1 and 2 Inspection At: Quad Cities Site, Cordova, IL Inspection Conducted: June 25 through August 19, 1989 Inspectors:
R. L. Higgins D. E. Jones R. M. Lerch M. Parker A. Dunlop J. M. Shine R. Bocanegra T. M. Ross F. J. Witt
/fA.bDwf Approved By:
R. M. Lerch, Acting Chief f-2f-E7 Reactor Projects Section 1B Date Inspection Summary Inspection on June 25 through August 19, 1989 (Reports No. 50-254/89016(DRP);
50-265/69016(DRP))
Areas Inspected: Routine, unannounced safety inspection by the resident, regional and Nuclear Reactor Regulation (NRR) inspectors of plant operations, radiological controls, maintenance / surveillance, licensee action on previous items, emergency preparedness, security, engineering / technical support and safety assessment / quality verification.
Results: During the inspection period one previously-identified unresolved item was determined to be a violation of NRC requirements (see attached Notice of Violation), and one new unresolved item was identified (see paragraph 8 of this report).
8909060200 89082s
$DR ADOCK 05000254 FDC
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DETAILS 1.
Personnel Contacted
- G. Spedl, Production' Superintendent
- R. Robey, Technical. Superintendent
- L. Petrie, Assistant Superintendent of Maintenance
- D. Gibson, Regulatory Assurance Supervisor l-
- J. Wethington, Quality Assurance Supervisor l
- T. Barber, Regulatory Assurance
- Denotes those,present at the exit interview on August 21, 1989..
The inspectors also contacted and interviewed other licensee and contractor personnel during the course of this inspection.
2.
Action on Previous Items (92702)
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a.
(Closed) Unresolved Item / Violation.254/89012-02; 265/89012-02:
Installation of Improper Check Valves. This item was reviewed
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and corrective actions described in paragraph 5.b(2) of Inspection
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Report 254/89012(DRP); 265/89012(DRP).
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As documented in the cover letter and enclosed Notice, this unresolved item was determined to be a Severity Level IV violation.
All corrective actions have been completed, therefore no response was required. This item is considered closed.
b.
(Closed) Violation 254/89012-01; 265/89012-01: Loss of Secondary Containment. This item was licensee-identified and all corrective actions to prevent a recurrence have been completed. This item is considered closed.
3.
Plant Operations a.
Operational Safety Verification (71707)
The inspectors, through direct observation, discussions with licensee personnel, and review of applicable records and togs, examined _ plant operations. The inspectors verified that all a
activities were accomplished in a timely manner using approved procedures and drawings and were inspected / reviewed as applicable; and that procedures, procedure revisions and routine reports were in accordance with Technical Specifications, regulatory guides, and industry codes or standards. Additionally, the inspectors verified that approvals were obtained prior to initiating any work; activities were accomplished by qualified personne); the limiting conditions.for operation were met during normal operation and while components or systems were removed from service; functional testing and/or calibrations were performed prior to returning components or systems to service; and independent verification of equipment lineup and review of test results were accomplished. Also verified were
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quality control records for being properly maintained and reviewed, and parts, materials and equipment for proper certification, calibration, storage, and maintenance as applicable. The inspectors l
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conducted frequent tours of plant facilities to search for the existence of adverse plant conditions such as equipment malfunctions,
I potential fire hazards, radiological hazards, fluid leaks, excessive vibrations, and personnel errors. The inspectors' review ensured these issues were addressed in a timely manner with sufficient and proper corrective actions and reviewed by appropriate management personnel.
During the inspection period a detailed plant tour was performed by the Branch Chief, Section Chiefs and Senior Resident Inspectors of the Region III Division of Reactor Projects Branch 1, by three of the Licensing Project Managers of Project Directorate III-2 of the Office of Nuclear Reactor Regulation, and by the Region III coordinator in the Office of the Executive Director for Operations.
ho significant safety discrepancies were found, but numerous minor deficiencies were noted., These were provided to the Senior Resident Inspector, who then presented them to licensee management. The licensee has instituted corrective actions to rectify these deficiencies.
b.
Engineered Safety Features System Walkdown (71710)
During plant tours of Units 1 and 2, the inspectors walked down some of the accessible portions of the High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), Core Spray (CS),
Residual Heat Ramoval (RHR), RHR Service Water, Standby Liquid Control (SLC) Systems, and Standby Gas Treatment (SGT) Systems. The inspectors also walked down the Emergency Diesel Generators (EDG)
and the Station Batteries. No violations or deviations were noted.
c.
Summary of Operations Unit 1 Unit 1 operated normally at the maximum power which fuel depletion would sustain (an operating status termed "coastdown"), or at reduced power in order to perform surveillance testing or respond to load dispatcher orders, until the reactor scrammed on June 29, 1989 due to a turbine trip caused by a spurious main condenser low vacuum signal (refer to paragraph 3.d.(1) of this report). Following
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maintenance work including replacement of the 1B recirculation pump seal, the reactor restarted on July 2, 1989, but shut down to repair a steam leak on a balance-of-plant component (refer to paragraph 3.d.(2) of this report). The reactor restarted on July 4, 1989, and oaerated normally at power in a coastdown status until July 7, 1989, w1en high down-river temperature limits established by the Illinois Environmental Protection Agency (EPA) restricted the reactor's power level (refer to paragraph 3.d.(3) of this report). The reactor power restrictions were removed on July 14, 1989, and the reactor operated in a coast down status for the remainder of the inspection period.
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Unit 2 i
Unit 2 operated normally at full power, on Economic Generation l.
Control (EGC), or at reduced power in order to perforu surveillance
testing or respond to load dispatcher orders, until July 7, 1989, I
when high down-river temperature limits established by the Illinois EPA restricted the reactor's power level (refer to paragraph 3.d.(3)
of'this report). The reactor power restrictions were removed on July 14, 1989, and the reactor resumed normal power operations.
On July 19, 1989, problems with the Core Monitorinq Code (CMC)
computer used to generate reactor power and critical heat flux limits malfunctioned (refer to paragraph 3.d.(4) of this report).
The reactor was maintained at a constant power level of 80% (650 MWe) until the malfunction was repaired on July 25, 1989. The reactor operated normally at power throughout the rest of the inspection period.
d.
Onsite Followup of Events at Operating Power Reactors (93702)
(1) Reactor Scram At 10:39 PM CDT on June 29, 1989, with Unit 1 at 100% power, the Unit 1 turbine tripped causing a reactor scram. Post scram actions were normal, except for a group 1 isolation (MSIV closure) which caused reactor pressure to rise resulting in an electromatic relief valve opening momentarily.
The group I isolation occurred because main steam header pressure dropped below the isolation setpoint pressure.
Immediately after the scram the licensee reset the group 1 isolation and placed the reactor in cold shutdown.
The cause of the June 29th turbine trip was determined to be a loose connection on the wire to the vacuum indicator 119it PS.105B on'the 901-7 panel. This loose connection would occasionally make and break electrical contact, generating a voltage spike on the vacuum indicator wire, which would in turn generate a voltage spike in the low vacuum turbine trip circuit, causing a turbine trip and reactor scram.
It is suspected that this problem also caused the Unit 1 turbine trip and reactor scram which occurred on December 5, 1988.
(Refer to paragraph 3.e.(1) of inspection report 50-254/88028(DRP)
and 50-265/88029(DRP)).
During the unplanned outage the licensee tightened the loose connection to PS 105B, replaced the malfunctioning seal on the 1B reactor recirculation pump, repaired the 1B reactor recirculation pump suction valve and performed other maintenance r ' vities.
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(2) Reactor Shutdown and Startup At 12:30 PM CDT on July 3, 1989, with Unit 1 at 25% power in the process of raising power following a reactor startup, the licensee discovered a steam leak on the 1A moisture separator drain tank vent line to the 1A moisture separator in the low pressure heater bay. The licensee shutdown the reactor at 10:17 PM on July 3, 1989, and entered cold shutdown at 3:50 AM on July 4, 1989.
During the unalanned maintenance outage the licensee repaired the steam leac on the moisture separator drain tank vent line, repacked the RCIC inboard steam supply valve and performed other minor maintenance in the drywell. At 3:49 PM CDT on July 4, 1989, the licensee began control rod withdrawal to restart Unit 1.
Reactor criticality was achieved at 6:16 PM on July 4, 1989, and the main generator was synchronized to the electrical grid on July 5, 1989.,
(3) Summer Dronht Operational Restrictions Quad Cities must abide by several environmental limits imposed by the Illinois EPA concerning the effect which circulating water discharging into the Mississippi River has on river temperature. These limits are:
(1) the maximum differential between up-river and down-river temperatures is not permitted to exceed 5 degrees F; (2) the maximum down-river temperature is limited to vabes which vary according to the month of the year (for the months of July and August this temperature limit is 86 degrees F), and down-river temperature is prohibited from exceeding these values for more than a total of 87 hours0.00101 days <br />0.0242 hours <br />1.438492e-4 weeks <br />3.31035e-5 months <br /> and 36 minutes in any 12 month period; (3) the down-rive. temperature is prohibited from ever exceeding the monthly temperature limit by more than 3 degrees F (for the months of July and August this temperature limit is 66 degrees F, never to exceed 89 degrees).
Because of drought conditions which existed during 1988, Quad Cities had only 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> remaining during the month of July
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before the time limit would be exceeded. On July 7, 1989, down-river temperature exceeded the limit for the first time in 1969. The limit was exceeded again on July 8, 9 and 10, 1989, as well. At 5:15 PM CDT on July 10, 1989, Qued Cities exceeded
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the 87 hour0.00101 days <br />0.0242 hours <br />1.438492e-4 weeks <br />3.31035e-5 months <br /> 36 minute time limit. The licensee immediately
reduced power on both units to 30% (250 MWe), and limited power i
to this 30% value through July 13. 1989, when the Illinois EPA I
granted Quad Cities an additional 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> during which the
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time-dependent temperature limit could be exceeded. This l
additional time, plus cooler weather, permitted Quad Cities to
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resume normal operations on July 14, 1989.
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(4) Core Monitoring Code Abnormality l
At 8:00 PM on July 19, 1989, with Unit 2 at 100% power, the reactor core thermal limits of MFLPD (maximum fraction of
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D generation rate) and MCPR (minimum critical power ratio) as.
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l-violation of-their limiting values. The licensee'promptly commenced a power reduction to 80% power.: Comparison with the-core thermal limits which were calculated 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> previously.
.and found to be within. limits, and the knowledge that no plant-
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-the licensee to question the validity of the core thermal limits which were calculated to. exceed their limiting values.
The. licensee continued power operation of Unit '.by holding p
p power constant at 80% and obtaining core thermal limits from B
corporate headquarters'and from. General Electric. The thermal L
limits so obtained were within their' limiting values.
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The abnormality in the core thermal limit calculation was determined to be due to a factor of ten error in a xenon constant for type'4 fuel in the core monitoring code. This l
problem was-insignificant at low fuel exposure but became more significant later in~ core life. The error was corrected, and Unit ~2 was returned to permal operaticn wit'h core thermal 111mits provided by the plant process computer on July 25, 1989.
(5) Loss of Service Air Compressor-On July 19, 1989, with both units operating at power, the
'2A service air compressor had to be shutdown because'it was spurting oil. The 1A service hir. compressor was out of service for maintenance. With only the 1B' service air-compressor'
operating, service air pressure began to decrease. Since the service air system was supplying the instrument air system,-
instrument air pressure'also began to' decrease. The cperating crew took action to minimize the use of service and instrument air, thereby arresting the pressure decrease, restoring service air and instrument air pressure, and preventing the possible shutdown of both units. The 1A service air compressor was subsequently returned to service and the service air;and instrument air systems returned-to normal operation.
(6) Degraded Reactor Recirculation Pump 2A Seal
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On August 10, 1989, operaticrs personnel noticed that the seal cavity pressure for the 2A reactor recirculation pump had increased from its normal value of approximately 500 psig at full power to 600 psig, indicating a degraded inner seal. The licensee promptly instituted a temporary procedure to monitor and rccord seal cavity pressure plus several other related parameters every four hours (this is the same temporary procedure as is described in paragraph 3.d.(10) in inspection report 254/89012; 265/89012). Seal degradation has been 0-gradual, and at the end of-the inspection period the seal cavity pressure has increased to 700 psig. The licensee intends to shutdown in the latter part of August to replace the degraded seal.
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Radiological Controls (71707)
The licensee continued to demonstrate noteworthy performance in the area'of radiological controls. Unanticipated maintenance activities including the replacement of the IB reactor recirculation pump seal were performed in the Unit 1 drywell wid no associated personnel contaminations. Personnel exposure was less than budgeted despite unant'cipated maintenance activities in the drywell and otner potentially high radiation areas. Several more personnel contaminations than budgeted occurred in the latter part of July, but only one personnel contamination occurred in the August portion of the reporting period.
During the inspection perlod the inspectors monitored the radiological survey of the low pressure heater bay and the transfer of resin from the Unit 2 phase separator to the Unit 1 phase separator. The licensee's performance, and the attention showed by management to further reduce personnel contaminations and exposure, is indicative of strong licensee support for the ALARA program.
No violations or deriations were noted.
5.
Maintenance / Surveillance a.
Monthly Maintenance Observation (62703)
Station maintenance activities of safety-related and nonsafety-related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with Technical Specifications.
The following items were considered during this review: the limiting conditions for operation were met while compor.ents or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable. Additional items reviewed included verification that functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; and activities were accomplished by qualified personnel. Also, the inspectors verified that parts and materials used were properly certified; radiological controls were implemented; ar.d fire prevention procedures were followed. Work requests were reviewed to determine the status of outstanding jobs and to assure that priority is assigned to the maintenance of safety related equipment which may affect system performance.
Portions of the following activities were observed / reviewed:
(1) Unit 2 Control Rod Drive Hydraulic Control Unit (54-27) Scram Inlet and Outlet Valves Overhaul and Inspection.
(2) Unit 2 Control Rod Drive Hydraulic Control Unit (54-27) Ball Check Yalve Inspection / Replacement.
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(3) Unit 2A Service Air Compressor Overhaul.
(4) Unit 1C Reactor Feed Pump Auxiliary 011 Pump Maintenance.
(5) Unit 1A. Moisture Separator Drain Tank Elbow Repair.
No violations or deviations were noted.
b.-
Monthly Surveillance Observation (61726)
The inspectors observed surveillance testing required tv the Technical Specification and verified that testing was performed
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in accordance with adequate procedures, that test instrumer.tation i
was calibrated, and that limiting conditions for operatior were met. Additionally, the inspectors observed / verified the removal and restoration of the affected components, and that test results conformed with Technical Specifications and procedure requirements.
Also. the inspectors verified that the results 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 menagement personnel.
Portions of the following activities were observed / reviewed.
i (1) Unit 2 Core Flux Mapping Using the Traveling Incore Probe.
(2) Local Leak Rate Test of the Unit 1 Personnel Hatch.
I (3) Unit 2 High Reactor Vessel. Pressure Scram Surveillance.
I (4) Unit 2 High Drywell Pressure Scram Surveillance.
(5)
1/2 B Standby Gas Treatment System Torque Switch Testing.
(6) Calibration checks of steam tunnel temperature s tectors.
6.
Emergency Preparedness (71707)
During the inspection period the inspectors irspected the Quad Cities Technical Support Center (TSC) and the Emergelcy Operations Facility (EOF).
No violations or deviations were noted.
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Security (71707)
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During the inspection period the inspectors toured the plant and the Central Alarm Station to assure that security programs were being properly implemented. The inspectors verified that security barriers were in place, security doors were operable, the security force was alert, personnel correctly displayed their identification badges and visitor access was being properly controlled.
No violations or deviations were noted.
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- a- ' Loss of Security Electrical Power
- At 8:48 AM CDT on June 28, 1989, electrical power was lost to the security computers, security television cameras, security alarms,
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and' central and secondary' alarm station consoles. The-licensee immediately initiated. compensatory me,asures, began to search for the-
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cause of the loss of electrical power, and took action to restore l
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- electrical power to the' security equipment.: At 9:32 AM the NRC'
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~ Emergency Operations Center was notified, and at 10:04 AM electrical-
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power was restored to the. security equipment.
The cause of the loss of electrical power to-the security equipment l
was the opening of the wrong circuit breaker while tagging out the
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l security diesel generator. When the breaker was reshut, electrical-j l
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power was restored.
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Security Drill l
l On July 15, 1989, the Senior Resident. Inspector monitored the
. performance of a security drill involving a simulated attack of I.
the plant and occupation of'several plant buildings by security personnel playing the parts of terrorists. The drill. involved-the participation of local law enforcement agencies and included u-simulated fires, hostages and casualties.
8. -- ' Engineering / Technical Support a.
'RHR Service Water Pump Discharge Pipino Modification (37828)
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During the inspection period the inspectors monitored the installation of the replacement piping for,the 2B RHR service water q
discharge pipe. The only problem noted was the apparent inadequate j.
support of compressed gas cylinders.. Several compressed gas-l cylinders were found to be held in their upright position by wire,
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I while several others at other plant locations were found to be held u
l-in their upright position by rope. Both the rope and wire methods.
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of. supporting compressed gas cylinders appear to violate the rules for the control of compressed gas cylinders stated in the Flammable l-and Combustible Material Control procedure QAP 1700-1. These I
apparent deficiencies were brought to the licensee's attention and i
corrected.
b.
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During the inspection perind the inspectors monitored the l
installation and modification testing of the feedwater hydrogen
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addition system.
The results of an inspection of the feedwater
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hydrogen addition system which was performed in May 1989 by NRC
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headquarters personnel were published during the inspection period.
The NRC headquarters personnel determined that the feedwater hydrogen addition system complied with the applicable Electric Power Research Institute guidelines.
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Neutron Radioassay Testing of the Unit 1<High Density Fuel Racks >
The Licensing Project Manager for Quad Cities (an inspector from.
'i-NRC headquarters) monitored a portion of the neutron radioassay test of the. Unit 1.high-density fuel. racks (HDFRs) which was performed by~the National Nuclea_r Corporation to determine the extend of Boraflex' degradation and gap growth ~.
-Neutron ~ attenuation tests (both the Standard and Special Test Methods) of the Boraflex-lining in Unit-1 HDFRs, as observed, was aerformed in a deliberate and controlled manner by experienced and knowledgeable personnel..Compliante with-the unique test
' procedure instructions and radiological controls was deemed-acceptable. The overa11' testing evolution was. accomplished consistent with previous licensee commitments. ' Appropriate Precautions were-taken!(i.e. redistribution of spent fuel in adjacent discharge racks) at the onset to preclude any potential
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-inadvertent criticality problems considering' introduction of a.
' Californium neutron source into the spent fuel pit.
The licensee is working in conjunction with the Electric Power Research Institute (EPRI) to develop an industry data base on-
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performance of Boraflex in HDFRs. _ Based upon the results of the
- analyzed test data gathered by the licensee and EPRI, gap grov:th
.versus exposure will be trended and evaluated to determine the validity of the " Gap Growth Nodel" as a bounding assumption for the spent fuel pit criticality analyses. This will allow.for-continued verification of Technical Specification 5.5.B (i.e.
less'than.95 Keff) compliance and reassessment of the surveillance-program testing intervals (presently conducted after each refuelina cycle).- The aforementioned results and data will be completed in'
several months. =The NRC' plans to review this information with the licensee at that time. ' ~
d.-
Requalification Training During the inspection period the inspectors monitored two separate qualification training exercises: a simulated fire in the cable spreading room on June 29, 1989, and a simulated plant shutdown from outside the control room vn July 11, 1989. Both training exercises were realistic, well-controlled by licensee training personnel, and beneficial for the participants.
e.
Configuration Control During the course of the inspection the inspectors noted several deficiencies in plant' drawings and several mislabeled valves.
Until the extent and safety significance of these deficiencies is determined, these deficiencies will be treated as an unresolved item (254/89016-01(DRP); 265/89016-01(DRP)).
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Safety Assessment / Quality Verification a. ' Evaluation of Licensee Quality Assurance Program Implementation U
(35502)
Dur:ng the inspection period the' Senior Resident Inspector observed quality assurance personnel inspecting portions of the 28 RHR service water pump discharge piping modification,'and attended the exit meeting for;the industrial hygiene audit performed by the
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Quality Assurance staff.
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iThe quality assurance inspectors were conscientious and,the audits were thorough and detailed.
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b..
In-Office Review of Written Reports of Nonroutine Events at Power Reactcr Facilities (90712) and Onsite Followup of Written Reports of-Nonroutine Events at Power Reactor Facilities) (92700)
During-the' inspection period the resident inspectors reviewed
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incidents such as scrams, ESF actuations and component failures which occurred at other plants..The resident inspectors informed the licensee of the details of all events which potentially had'
applicability to components or activities at Quad Cities.
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LER Review (1)
(Closed) LER 254/87006 Revision 01:
Unit 1.HPCI Inoperable Due to Loose Solenoid Soldered Connection.
-This revision updates a previous LER by documenting the completion of the long term corrective actions taken to ensure-that similar events are avoided.
This item is considered closed.
(2)
(0 pen).LER 254/87017, Revision 00: High Pressure Coolant Injection System Inoperable due to Invalid System Isolation from Failed Differential Pressure Switch.
Licensee corrective ac.lons for this event are partially complete. This LER will remain open until the supplemental report is issued.
(3)- (Closed) LER 254/87025 Revision 00:
Control Room Habitability Design Basis Assumption Error Results in Exceeding Allowable Filter Efficiency.
This item was discussed in Inspection Report 254/87033(DRP).
Tracking of this item has been assumed by Region based-inspectors under 254/87025-L2. This intem is closed.
(4)
(0 pen) LER 265/89003 Revision 00: Loss of Secondary Containment Integrity.
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This event is discussed in paragraph 3.d.(6) of Inspection
. Report 254/89012 and 265/89012.
It will remain open pending revision of Q0A 6900-2, Loss of Unit One 125 VDC Supply, and Q0A 6900-4, Loss of Unit Two 125 VDC Supply, in accordance with the corrective actions stated in the LER.
(5)
(0 pen) 254/89004 Revision 00: Manual Reactor Scram Due to Relief Valve 1-203-3D Stuck Open.
This event is discussed in paragraph 3.d.(7) of Inspection Report 254/89010. This item will remain open pending the revision of the Electromatic Relief Valve (ERV) Disassembly, Inspection, Repair and Reassembly procedure, QMMP 203-22, to include a check off step to verify that the drain orifice in the dire retainer of the main valve is clear.
It will also remain open until the licensee receives the report from Dresser Industries and addresses all additional actions which Dresser recommends be implemented to enhance the reliability of the j
ERVs and pilot valves.
(6)
(Closed) LER 254/89006 Revision 00: Recirculation Pump Trip.
This is a voluntary LER which covers an event discussed in paragraph 3.d.(9) of inspection report 254/89012 and 265/89012.
All corrective actions are complete. This item is considered clcsed.
(7)
(Closed) LER 254/89007 Revision 00:
lA Refuel Floor Radiation Monitor Spiked high.
i This event is discussad in p,.ragraph 4.c. of Inspection Report 254/89012 and 265/89012. All corrective actions have been
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completed. This item is considered closed.
(8)
(Closed) LER 254/89008 Revision 00:
Inability of Pressure Valve 1-1601-23 to Fail Safe.
l This event is discussed in paragraph 5.b.(2) of Inspection Report 254/89012 and 265/89012.
It will remain open pending revision of QOS 1600-15, Pressure Suppression System
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Power-0perated Valve Quarterly Fail-Safe Test, to test for both slow and fast loss of instrument air.
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(9)
(0 pen) LER 254/89009 Revision 00: Unmonitored Release of Laundry Water to the Discharge Bay.
This event is discussed in paragraph 4.b. of Inspection Report l
254/89012 and 265/89012.
It will remain open pending the l
replacement of the skid drain line piping now in use with I
piping designed for acidic / caustic applications.
(10) (Closed) LER 254/89010 Revision 00: Reactor Scram Due to a Loose Wire on the Condenser Low Vacuum Pressure Switch Indicating Lamp.
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iThis event is discussed in paragra n 3.c. of'this report..
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.All.short term corrective actions lave been completed, and
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La1171ong-term corrective actions are being tracked by the l11censee's-Nuclear Tracking System. This item is considered-
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'(11)'(Open) LER 254/89011 Revision 00:
Unit 1 Diesel Generator Fire-Protection System Operable.
During surveillance testing of the Unit 1 diesel generator on July.7, 1989,'the normal vent supply. damper was found open when it should have been closed. The cause was determined to be a:
pipe plug which had been left in the exhaust port of the' damper.
solenoid valve. 1The damper ~had been'. inoperable since July 14 1988, when the solenoid valve had been replaced and no operability test had been conducted. The plug was removed, the-system was tested satisfactorily and returned to service on July 8, 1989.
This' item will remain open pending the' completion of the corrective actions listed in the LER.
(12) (Closed) LER 265/87009, Revision 00: Scram Caused by Turbine /
Generator Load Mismatch Due to a Main Transformer "C" Phase Fault.
The licensee corrective actions reviewed by the resident-inspector appeared adequate. This item is closed.
c.
Evaluation of Licensee Self-Assessment Capability (40500)
-(1)~'On-Site Review Committee During the inspection period.the Senior Resident Inspector
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attended an On-Site Review Committee meeting on the' loss of I
security electrical power (refer to paragraph 7.a of this
.eport). The committee was properly staffed, adequately l
addressed the relevant issues, and demonstrated adequate concern for reactor safety.
(2) Licensee Self-Assessment During the inspection period the Senior Resident Inspector attended the exit meeting conducted by the licensee's operations self-assessment team. The assessment appeared thorough.and detailed, and the assessment team's findings l
were communicated in a professional manner.
No violations or deviations were noted.
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10. Management Meetings - Entrance and Exit Interviews (30703)
The inspectors met with licensee representatives (denoted in Paragraph 1)
throughout the inspection period and at the conclusion of the inspection on August 21, 1989, and summarized the scope and findings of the inspection activities.
The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the ins)ectors during the inspection. The licensee did not identify any suci documents / processes as proprietary.
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