IR 05000269/1989040
| ML15224A629 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/07/1990 |
| From: | Shymlock M, Skinner P, Wert L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML15224A628 | List: |
| References | |
| 50-269-89-40, 50-270-89-40, 50-287-89-40, NUDOCS 9002230541 | |
| Download: ML15224A629 (15) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, ATLANTA, GEORGIA 30323 Report Nos: 50-269/89-40, 50-270/89-40, 50-287/89-40 Licensee: Duke Power Company 422 South Church Street Charlotte, N.C. 28242 Docket Nos.: 50-269, 50-270, 50-287 License Nos.: DPR-38, DPR-47, DPR-55 Facility Name: Oconee Nuclear Station Inspection CQnducted: December 17, 1989 - January 13, 1990 Inspectors:'~ )
/,A
~
Z~
-7~
P. H. Skinner/, enior Residdt Inspector Date Signed L. D. Wert, Resdent Inspector Date Signed Approved by:
'
-)
M. B. Shymlo6k, Chief Date Signed Reactor Projects Section 3A Division of Reactor Projects SUMMARY Scope:
This routine, announced inspection involved inspection on-site in the areas of operations, surveillance testing, maintenance activities, non-routine reporting program, installation and testing of modifi cations, and inspection of open item Results: One example of recently cited Violation 50-269,270,287/89-36-02:
Failure to Follow Procedures During Operation in a Shutdown and Cooled Down Condition, was noted during this period (paragraph 2.c).
A weakness was noted concerning an apparent lack of attention to detail and Operations personnel knowledge regarding the operability of the Emergency Condenser Circulating Water System (paragraph 2.d).
In addition to the routine inspection activities, the residents reviewed the licensee's actions concerning:
-
Operation of Units 1 and 2 with Reactor Coolant Pump Monitor setpoints not as required by TS. (paragraph 2.b)
-
A series of problems concerning the operability of the Unit 3 Turbine Driven Emergency Feedwater Pump (paragraph 4.b).
REPORT DETAILS 1. Persons Contacted Licensee Employees
- B. Barron, Station Manager D. Couch, Keowee Hydrostation Manager
- J. Davis, Technical Services Superintendent D. Deatherage, Operations Support Manager W. Foster, Maintenance Superintendent T. Glenn, Instrument and Electrical Support Engineer D. Hubbard, Performance Engineer
- E. Legette. Compliance Engineer
- H. Lowery, Chairman, Oconee Safety Review Group
- B. Millsap, Maintenance Engineer D. Powell, Station Services Superintendent
- G. Rothenberger, Integrated Scheduling Superinteneent R. Sweigart, Operations Superintendent Other licensee employees contacted included technicians, operators, mechanics, security force members, and staff engineer MlRC Resident Inspectors:
- P Skinner
- L. Wert
- B. Desai
- Attended exit intervie. Plant Operations (71707)(71710)(71711)
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, shift turnover records, and equipment removal and restoration records were reviewed routinel Discussions were conducted with plant operations, maintenance, chemistry, health physics, instrument &
electrical E), and performance personne Activities within the control rooms were monitored on an almost daily basi Inspections were conducted on day and on night shifts, during week days and on weekends. Some inspections were made during shift change in order to evaluate shift turnover performanc Actions observed were conducted as required by the Licensee's Administrative Procedure The complement of licensed personnel on each shift inspected met or exceeded the requirements of T on Operators were responsive to plant annunciator alarms and were cognizant of plant condition Plant tours were taken throughout the reporting period on a routine basis., The areas toured included the following:
Turbine Building Auxiliary Building Units 1, 2 and 3 Electrical Equipment Rooms Units 1, 2 and 3 Cable Spreading Rooms Units 1, 2 and 3 Penetration Rooms Station Yard Zone within the Protected Area Standby Shutdown Facility Units 1, 2 and 3 Spent Fuel Pool Rooms Keowee Hydro Station During the plant tours, ongoing activities, housekeeping, security, equipment status, and radiation control practices were observe Unit 1 -
Unit 1 operated at 100 percent power for the entire report period with the exception of several short periods at reduced levels for load following purposes and one interval at 86 percent for performing main steam stop and control valve functional testin Unit 2 -
Unit 2 operated at 100 percent power for the entire reporting perio Unit 3 -
Unit 3 was brought critical on December 18 for zero power physics testing following the completion of end of cycle 11 refueling outag When attempts were made to place the generator on the line on December 20, a sheared wire was identified on the exciter rotor (supplies generator field excitation) circuitr Repairs to the exciter rotor, to several nuclear instruments that had also been found defective, and to the Turbine Driven Emergency Feedwater Pump were completed and the unit was brought on-line December 2 A problem with one phase of the generator output motor operated disconnect delayed power escalation for several hour Power was then increased to 93 percent and 99 percent while adjusting the steam generator water level control on the 3B generator. Apparently fouling of the broached tube support openings had resulted in increased steam generator water levels as power was increased. This problem previously occurred on Units 1 and 2 and was resolved by chemical cleaning (see Inspection Reports 50-269,270,287/87-40 and 50-269/84-27, 50-270/84-23, 50-287/84-25).
The unit reached 100 percent power on December 28. With the exception of a short period at 95 percent to correct a steam leak on a heater drain valve, the unit remained at this power for the remainder of the reporting perio The licensee is presently considering longterm corrective actions to resolve the steam generator high level condition During the inspection period the inspector participated in an Emergency Preparedness (EP) drill conducted on January 10, 1990. The drill was a simulated steam generator tube leak and loss of offsite power casualty on Unit 2. The drill was designed to be a table top drill and only the Technical Support Center and the Operational Support Center personnel were participating. Communications and dose assessment equipment were exercise The drill provided valuable training for alternate personnel that may be required to fulfill key personnel positions and enabled the recently reported Station Manager to adjust to his role as Emergency Coordinato The inspectors provided feedback on observations during the drill to station management and the station EP manager. The inspectors will continue their involvement in onsite drill b. Operation of Units 1 and 2 With Reactor Coolant Pump Monitor Setpoints Not As Required by Technical Specification At about 1 on December 29, 1989, the licensee informed the resident inspector that a problem had been discovered regarding Technical Specification (TS) 2.3:
Limiting Safety System Settings, Protective Instrumentation. Amendment No. 180, 180 and 177 to Units 1, 2 and 3 respectively was issued on December 15, 198 The amendments principally addressed the most recent Unit 3 core reload requirement It also included the use of a different thermal hydraulic code and the modification of TS associated with power operations with only two Reactor Coolant Pumps (RCP)
operatin Prior to these amendments, TS 2.3 required the RCP monitors to produce a reactor trip for the following conditions:
-loss of two RCPs and power level greater than 55 percent of rated power-loss of two RCPs in one loop and power level greater than 0 percent of rated power-loss of one or two pumps during two RCP operatio This amendment requires the pump monitor instrumentation to produce a reactor trip when a loss of two pumps occurs and power level is greater than zero percent power. Apparently the reason this change was made was to avoid the cost required to obtain an analysis of the reload configuration with two RCP operatio In order to promote consistency between the three units, the limit was changed on all units although Unit 3 core reload is the only core currently not analyzed for this operatio On Unit 3, Instrumentation and Electrical (I&E)
personnel had adjusted the setpoints on the pump monitors to meet the current TS requirements but, due to a communication problem, the adjustments
were not made on Units 1 and 2. The licensee stated that extensive procedure revisions would have to be completed prior to making this adjustment, and it was not a frequently performed evolution. Since the Unit 3 setpoints (the only setpoints that core reload actually effected) had been reset and Units 1 and 2 are analyzed for 2 RCP operation at less than 55 percent power, the licensee identified that these setpoints would not be adjusted until about January 5, 199 The details of this decision were discussed with the resident inspector. The licensee also identified the following guidance that would be provided to the operator in the interim period:
-
On Units 1 and 2 the RCP monitors are considered conditionally operable when reactor power is greater than 55 percen If Unit 1 or 2 is operated at less than 55 percent, then the monitors are considered inoperabl TS Limiting Condition for Operation (LCO)
3.0 would be entered for that unit and one of the following actions must be completed within the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> action statement:
-
change the monitor setpoints to meet the TS requirements within the prescribed time limits and exit the LCO (or place the unit in hot shutdown)
-
increase power to greater than 55 percen If two RCPs are lost orn Unit 1 or 2, that unit will be shutdown in an orderly manger within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, in accordance with TS (If the unit is operating at greater than 55 percent a trip will occur.)
Discussions were held with the Oconee NRR Project Manager and regional staf Since there is no safety concern associated with operating in this manner, regional staff concurred with the above actions. Regional staff requested the licensee to submit a letter to NRC Region II fully documenting this issue and the corrective actions. This letter was submitted as requested (See DPC letter to Mr. S. D. Ebneter, Regional Administrator dated January 3, 1990) with a commitment completion date of January 8. The monitor setpoints were changed on Unit 1 on January 3, and Unit 2 setpoints were changed January 4. A report will be submitted in accordance with 50.73(a)(2)(i)(B).
c. Misposition of 3LP-1 Breaker On December 1,
during a tour of Unit 3 plant areas, the inspector noted that the power supply breaker for valve 3LP-1, decay heat removal supply valve, was not in the position prescribed by Operating Procedure.(OP) 3/A/1104/04, Low Pressure Injection Syste This breaker is specified by this OP to be open and independently verified as open under the existing plant conditions. In addition, there is a descriptive label on the breaker panel that also states the breaker must be open when primary system pressure is greater than 350 psig to meet 10 CFR 50 Appendix R criteria. Investigation by the licensee identified that the OP to position this breaker had been correctly performe However, subsequent work on the valve was performed and testing accomplished by I&E procedure IP/O/A/3001/2, Setting of Limitorque Switches on Rotork Valve Operator The valve breaker final position was not addressed in the I The unit supervisor (US) directed the non-licensed operator (NLO) to leave the breaker shut without any verification that this position was correc The NLO did not discuss with the US that the label on the breaker specified that the breaker be left in the open positio OP/0/A/1102/06, Removal and Restoration of Station Equipment, dated November 7, 1989, states that one of the purposes of this procedure is the removal from service or change from a procedure designated status of station equipmen This failure to follow procedure OP/0/A/1102/06, is identified as another example of procedure adherence deficiencies previously cited as Violation 50-269,270,287/89-36-0 The licensee will respond to this example as a part of the response to Violation 89-36-0 Unit 3 Emergency Condenser Circulating Water System Inoperability On December 19, 1989, during a walkdown of the Unit 3 Control Room indications, the inspector noted that valve 3CCW-26, a' high point automatic vent valve on the +/-3A'
Condenser Circulating Water (CCW)
intake header indicated partially open. This valve serves as a high point vent during normal operation of the CCW system. The CCW system is designed such that on a loss of power situation (loss of CCW pumps), gravity flow and a siphon effect will cause continued flow of CCW through the pipin This portion of the system is called the Emergency Condenser Cooling Water (ECCW)
syste During such operation a vacuum is maintained by steam air ejectors in the piping to sustain siphon flow from the lake level through the pump up to the intake piping leve If the CCW intake vent valve was left open this could possibly prevent the siphon effect from functioning and result in loss of ECCW flo The ECCW system is required to be operable by Technical Specification 3.4.5.a whenever the reactor is above 250 degrees Discussions with the control room operators indicated that some personnel did not understand the consequences of this valve being mispositioned. ECCW was not declared inoperable and the applicable 7, day LCO had not been entere An operator was dispatched to the valve and verified it was partially open. After discussion involving the inspector, the operators and the shift manager, the ECCW system
was declared inoperable effective from the previous da The CCW pumps had been cycled that day which caused operation of the vent valves and 3CCW-26 had apparently failed to fully shu The valve was subsequently fully shut to restore the system to an operable statu Inspection Report 269.270,287/89-12 discusses a similar situation concerning maintenance on the operator of high point vent valve 3CCW-28. At that time the inspectors questioned the operations staff regarding control over the position of these valves. Included in the licensee's prompt response to the inspectors concerns was a letter issued by the designated CCW system experts addressing ECCW operability which included a discussion of vent valves. This letter (dated May 1, 1989) is in the active Operations Guide Notebook which provides guidance to the operators and specifically states if a high point vent valve is open for an extended period of time, the ECCW system must be declared inoperable. This apparent lack of attention to detail regarding the operability of a TS required system and the apparent lack of knowledge concerning these valves is identified as a weaknes No violations or deviations were identifie. Surveillance Testing (61726)
Surveillance tests were reviewed by the inspectors to verify procedural and performance adequacy. The completed tests reviewed were examined for necessary test prerequisites, instructions, acceptance criteria, technical content, authorization to begin work, data collection, independent verification where required, handling of deficiencies noted, and review of completed work. The tests witnessed, in whole or in part, were inspected to determine that approved procedures were available, test equipment was calibrated, prerequisites were met, tests were conducted according to procedure, test results were acceptable and systems restoration was complete Surveillances reviewed and witnessed in whole or in part:
PT/O/A/0150/22K Condenser Circulating Water Valve Functional Test, dated 12/12/88 (3CCW93)
IP/2/A/0305/001 Reactor Protective System Channel Pump Power Monitor Instrument Calibration PT/3/A/0600/13 Motor Driven Emergency Feedwater Pump Performance Test dated 11/29/88 PT/3/A/0600/12 Turbine Driven Emergency Feedwater Pump Performance Test dated 12/20/89 PT/0/A/0160/06 Reactor Building Cooling Unit Heat Exchanger Performance Test dated 2/21/89 No violations or deviations were identifie. Maintenance Activities (62703)
a. Maintenance activities were observed and/or reviewed during the reporting period to verify that work was performed by oualified personnel and that approved procedures in use adequately described work that was not within the skill of the trad Activities, procedures, and work requests were examined to verify; proper authorization to begin work, provisions for fire, cleanliness, and exposure control, proper return of equipment to service, and that limiting conditions for operation were me Maintenance reviewed and witnessed in whole or in part:
WR 25564C Unit 2 Repair Valve 2CF-3 WR 25405C Unit 3 Turbine Driven Emergency Feedwater Pump Turbine Governor Control Valve Work (MP/O/A/1200/012)
WR 054514 Hydrostatic Testing of the +/-2B' Low Pressure Injection System WR 25411C 3MS-87 Not Controlling Steam to Turbine Driven Auxiliary Feedwater Pump Turbine Properly WR 25547 High Pressure Service Water Jockey Pump Packing Leakage WR 25801C Replace Nitrogen Bottles for 3MS-87 b. Unit 3 Turbine Driven Emergency Feedwater Pump Problems During the startup of Unit 3 following the refueling outage, several operational problems were identified and subsequently corrected concerning the Turbine Drive Emergency Feedwater Pump (TDEFWP):
-
Testing of the TDEFWP on December 14, indicated that valve 3MS-95, the steam governing valve for the TDEFWP turbine did not fully shut after a test run. The apparent cause was that-small pieces of slag material had become stuck in 3MS-95, preventing it from functioning properly. The lower portion of the control and trip/throttle valve (specifically the control or governing portion) was inspected and cleared of foreign material, the pump was operated and the valve again inspected/cleared. Earlier in the outage check valves had been replaced in both the main steam (MS)
and auxiliary steam supply lines to the TDEFW Apparently, as a result of poor maintenance practices, this resulted in welding slag material being present in these steam lines following the maintenance. On December 17, the TDEFWP was declared operable after PT/3/A/0600/12 (TDEFWP Performance Test)
was satisfactorily complete On December 20 the TDEFWP was removed from service for repacking of the bearing of oil cooling pum The maintenance was completed and the pump was returned to service after satisfactory completion of PT/3/A/0600/1 At that time the unit was proceeding back to hot shutdown from low power range operation to work on Nuclear Instrumentation and a control rod position indication proble On December 21 the TDEFWP was again declared inoperable due to possible overpressurization of the EFWP discharge pipin During the performance of PT/O/A/0600/18:
EFW Train Operability Test, on December 15, 1989, some data had been collected as required for post maintenance testing. Apparently one operator had recorded that a local pump discharge pressure gauge indicated 2000 psig. A review of the testing documentation by Quality Control personnel identified that this was an abnormally high discharge pressur It was suspected that the MS-95 problem had allowed excess steam flow to the turbine during the testing. A calculation indicated that if the turbine had run up to its overspeed setpoint (it tripped once during testing on the December 15-17 period but it is not sure if overspeed was the cause) the maximum pressure the discharge piping would have been exposed to would be 2519 psi The licensee's design engineering group completed an operability evaluation stating that the turbine, pump, valves, piping, and fittings were all considered operable. Code allowable hydrostatic test pressures had not been exceeded on the piping. Visual inspections were conducted at cperating pressure. Instrumentation which may have been affected was recalibrated. At about midnight on December 21. the TDEFWP was run in recirculation for training in accordance with a modified procedure. OP/3/A/1106/06-TDEFWP, to ensure no damage had occurred. At the end of this test it was identified that valve 3MS-95 had again stuck partially ope The TDEFWP remained inoperable. Although other activities were nearing completion which would allow the unit to return to operation, station management decided that Unit 3 would not be brought critical until the TDEFWP problem had been resolve (Oconee's TS do not restrict mode changes with inoperable equipment.) After some additional effort, maintenance personnel removed the upper portion of the trip/throttle valve which contained a strainer assembly designed to prevent material large enough to cause turbine damage from entering the turbine (this portion had previously not been disassembled). Larger pieces of slag material and some other material was found lodged against this scree The screen holes were about an eighth inch in diameter which apparently permitted the passage of pieces large enough to affect the operation of the governor valve portion of MS-95. At about 9:30 the reactor was brought critica The TDEFWP was tested periodically after unit startup to ensure no further problems had occurre During some of the testing on December 22, it was discovered that.valve 3MS-87 (main steam supply to TDEFWP pressure regulating valve) had a significant seat leak and had caused the MS supply piping between 3MS-88, 3MS-89 and 3MS-93 to be overpressurized to 900 psi The problem was identified by observing an overranged pressure transmitter in the steam lin The MS supply line to the TDEFWP was isolated by shutting #MS-82 and 3MS-8 The TDEFWP was separated from auxiliary stea Repairs were completed to valve 3MS-87 (the internals were replaced)
and the pressure transmitter was repaire The MS supply to the TDEFWP was declared operabl A subsequent operability evaluation of the MS line overpressurization stated that all the concerned piping and valves are capable of performing their intended safety function and could be returned to service without corrective actio The resident inspectors followed the licensee's actions throughout the above effort With the exception of the apparent error which caused the material to get into the steam lines and a delay in inspecting the screen on the upper portion of the trip/throttle valve, the actions appeared appropriate and conservative. Attention to detail enabled the licensee to identify the problem with 3MS-95 during earlier testing. Actions taken following the identification of the potential overpressurization were prompt and appeared thoroug Discussions with maintenance management indicate that in the future
"flame cutting" (technique used to remove the steam line check valves which had resulted in the 3MS-95 problems) would not be utilized in similar configuration Some of the other material found in the 3MS-95 screen was attributed to an earlier failure of a valve in the Auxiliary Steam syste Discussions were held with station management regarding the their decision to not inspect the screens on the Unit One and Unit Two TDEFWP governor valves for foreign materia Inadvertent Actuation of RPS and ES Channels (Unit 2)
On January 10, 1990, at 12:58 an inadvertent actuation of Channel A Reactor Protection System (RPS)
and Emergency Safeguards (ES)
channel A occurre The actuation caused the following problems:
-
All trip bistables on channel A tripped which also caused the opening of one of the control rod drive breaker ES channel A tripped resulting in annunciator alarms on various ES equipmen A momentary loss of train A of the Inadequate Core Cooling Monitoring Syste The above equipment was rapidly returned to norma The investigation into this problem identified that a chart recorder used to provide trending for selected radiation process monitors had been repaired and was being reinstalled into its panel in the Unit 2 control roo During the installation process, a short occurred while sliding the recorder into the housing which in turn caused a voltage spike on KVIA power suppl This power supply is also the power source for RPS and ES channels The recorder was disconnected and removed from service. It has been taken to the I&E shop for further troubleshootin The inspectors will continue to follow the licensee's actions regarding this even No violations or deviations were identifie. Nonroutine Reporting Program (90714)
The inspector reviewed the licensee's administrative control program for review, evaluation and reporting of nonroutine events or issue The following documents were reviewed in detail:
-
Administrative Policy Manual (APM)
Revision 28, Section 2.8, Reporting
-
4.5.5, Problem Investigation Process, dated July 24, 1989 Nuclear Production Department Directive (NPDD), Problem Investigation Process, Revision 5
-
ONS Compliance Manual Procedure 4.7, Processing Non-Conforming Item reports, dated April 8. 1986
-
ONS Compliance Manual Procedure 4.9, Problem Investigation Reporting, dated August 15, 1989
-
NPDD 4.8.1, Revision 1, Operating Experience Program Description The documents identified above provide details on the methods to be used to disseminate, review, take corrective actions, and report safety-related events occurring at the site. The Problem Investigation Report (PIR) is the primary mechanism for accomplishing these activitie Copies of all PIR's generated are routed to the inspectors for their information. The licensee's procedures also address in detail, actions to be taken upon receipt of vendor (including NRC) bulletins and circular PIR's were reviewed in detail during an inspection conducted by regional based inspector Refer to Inspection Report 50-269,270,287/90-01 for additional information addressing this subjec No violations or deviations were identifie.. Installation and Testing of Modifications (37828)(37701) (Unit 3)
The inspectors reviewed portions of several completed Nuclear Station Modification (NSM)
package Emphasis was placed on ensuring work was completed in accordance with the NSM and that post installation testing was completed as require Inspection Report 50-269,270,287/89-36 contains additional discussion concerning these inspection module The installation of cabling for NSM 32803: Main Feeder Bus (MFB)
Safety Train Cable Separation, was walked dow No discrepancies were identifie Installation procedure TN/3/A/2803/0/00:
NSM 2803: Main Feeder Bus Cable Separation, was reviewed along with the required maintenance procedure The inspector noted that the 10CFR50.59 analysis and the TN identified +/-cold shutdown' as initial condition Since the installation of this NSM required MFB number two to be de-energize additional restrictions of core defueled should have been included in the T All installation work was done in a defueled conditio Conversa tions with projects personnel involved with the NSM indicates this was assumed even though not specifically stated in the procedure. During this inspection report period, regional based inspectors also examined several completed NSM packages including retesting of installed modification Based on the residents efforts and the inspection discussed in Inspection Report 50-269,270,287/90-01, this inspection effort is complete. Inspection of Open Items (92700)(90712)(92701)
The following open items were reviewed using licensee reports, inspection, record review, and discussions with licensee personnel, as appropriate: (Closed) Violation 50-287/88-28-02:
Inadequate Testing Procedure Resulting In A Violation of TS 3.8.3. The licensee responded to this violation in correspondence dated October 28, 1988 with a supplemental response dated December 1, 198 The actions taken have been completed by the licensee and the inspectors have monitored similar activities to assure that the actions associated with operations pre-job briefings have occurre Based on this review this item is close b. (Closed) Violation 50-269,270,287/88-35-01:
Inoperability of RBCU Dropout Plates. This violation was cited as a result of these plates failing to operate as required during functional testing. An NSM was completed which revised the plate design. Functional testing of the revised plates was observed on Unit 1 (See
Inspection Report
269,270,287/89-05).
Installation of the NSM was completed on Unit 3
during the End of Cycle 11 refueling outage in December 198 The
modification of the plates along with the regular scheduled
preventive maintenance inspections should prevent any reoccurrence of
plate inoperability. This item is close (Closed) LIV 50-269,270,287/88-08-07: Lack of Adequate Procedure for
Isolation of Equipment Containing EPSL (Emergency
Power Switching
Logic) Equipmen Operating Procedure (OP)/O/A/1107/11, Removal and
Restoration of Auxiliary Electrical Systems, dated November 21, 1989,
has been generated to address removal of equipment associated with
the EPSL syste Based on the issuance of this O this item is
close (Closed) IFI 50-269,270,287/88-08-01: Improvement in Cable Room Fire
Detection Systems. This item addressed a concern that the cable room
fire detection system may not be sufficiently sensitiv A small
fire had occurred in a Unit 2 non-safety related computer cabine Despite light to medium smoke, a smoke detector located only five
feet from the smoldering terminal board did not alar Subsequent
testing indicated it was properly functioning and the system was
fully operabl The licensee is currently designing a Nuclear
Station Modification (NSM ON-52795) which is intended to
substantially enhance the existing fire detection system. The design
phase of this NSM is scheduled to be completed in June of 1990. The
installation of most of the detectors is scheduled by the end of
199 Based on the licensee's documented intentions to improve the
fire detection system, this item is close (Closed) IFI 50-269,270,287/88-08-03:
Reactor Building Cooling Unit
(RBCU) Dropout Plate Inspection. This item documented a concern that
no requirement existed to verify the function of the RBCU dropout
plate The plates appeared to receive minimum preventive
maintenance attention despite their safety functio Maintenance
Procedure MP/0/A/3009/14:
RBCU-Fusible
Patches -
Preventive
Maintenance Inspection, was developed to address RBCU dropout plate
inspection. Subsequent functional testing of the plates resulted in
the plate design being revised. MP/0/A/3009/14 is being revised to
require an inspection of the plates each refueling outage in addition
to an actual "drop test" of one of the three plate This item is
close (Closed)
IFI 50-287/88-34-02:
Integrated Control
System (ICS)
Performance During Main Turbine Trip Runback. This item addressed
concerns that the ICS permitted feedwater flow to decrease much
faster and to a lower value than reactor power following a Unit 3
runback during a turbine trip and as a result a reactor trip
occurre A Station Problem Report (SPR-2570)
was submitted in
November 1988 addressing an ICS modification to improve system
functioning on a turbine trip or load shed event. The modification
would block the steam pressure correction to the Steam
Generator/Reactor Master demand during certain conditions. The SPR
is still active with the implementation of the modification currently
under evaluation. This item is close (Closed)
IFI 50-269,270,287/89-17-03:
Adequacy of Fire Protection
Sprinkler System Pressur This item was identified by the licensee
and reported in LER 269/87-11 dated January 4, 198 An evaluation
has been conducted and resulted in changes to the pre-fire pla These changes require the startup of a high pressure service water
(HPSW)
pump upon indication of a fire in the cable and equipment
roo Pre-fire Plan Manual Plan #8, Section 4.d and Section 5
contain instructions for the fire brigade leader to start a HPSW pump
as part of the immediate action. Based on this action, this item is
close (Closed) LER 269/89-02:
Fire in 1TA Switchgear Due to Unknown Caus This LER addressed a fire and subsequent reactor trip on January.
An Augmented Inspection Team (AIT) had closely examined the
circumstances of the event. Many of the corrective actions addressed
several violations which were cited as a result of AIT concern Included in the completed corrective actions was the addition of
approximately ten supplemental fire brigade members to each shif These shift maintenance, radwaste and health physics personnel have
completed initial training, are fully qualified fire brigade members
and are available to supplement the normal shift fire brigade
personne All other corrective actions are completed or in
progress. This item is close i. (Closed)
LER 287/88-06:
Reactor Trips Due to Unknown Cause and
Eauipment Failure. This LER was submitted by licensee correspondence
dated January 13, 1989. The corrective action identified in this LER
has been reviewed by the inspectors. Based on this review, this item
is close (Open)
LER 287/89-06:
Polar Crane TS Violated Due to Management
Deficiency, Inadequate Policy. The inspectors reviewed this LER and
had the following concerns:
-
The LER did not address all the sections cf TS 3.12 (Reactor
Building Polar Crane and Auxiliary Hoist) that were violate The safety analysis referred to Section 15.11.2.1 of the Final
Safety Analysis Report
(FSAR)
which analyzes a single fuel
assembly acciden The inspectors questioned the appropri
ateness of this reference since it appears that more than one
assembly may be damaged if the main hook or a load fell into the
vesse These concerns were discussed with members of the Oconee Safety
Review Group and station management. The licensee stated that the
LER will be supplemented. This LER will remain open pending review
of the supplemen II
14 (Closed) 1OCFR21:
Brown Boveri K-Line Circuit Breakers(P2189-01).
The licensee has received the Part 21 information from ASEA Brown
Boveri (ABB). A review of records by the licensee indicate that the
site has a total of 176 breakers that do not have the rebound spring
addressed by this notic An analysis was performed by the Design
Engineering group and concurred with ABB, that continued operation of
the breakers concerned should not be a proble The licensee has
ordered this part and is installing the springs during the routine
maintenance being conducted on the breakers during each units
refueling outage. Based on this action, this item is close.
(Closed) 10CFR21:
Information on B&W Plugs Fabricated From Heat W
592-1 (P2188-06). This Part 21 has been addressed also in Inspection
Report 50-269,270,287/89-28. The licensee has developed a program to
inspect plugs that are susceptible to stress corrosion crackin This program included plugs fabricated from heat W592-1 and other
susceptible heats identified by
B& This program consist of
inspecting 100 percent of all hot leg plugs and at least 20 percent
of cold leg plug As a result of the inspection of the steam
generators during the most recent outage on Unit 3, four plugs on the
A steam generator were replaced and no plugs were found in the B
steam generator that required replacemen Based on the
implementation of this program, this item is close (Closed)
10CFR21:
Discrepancies in Weights of Limitorque Valve
Operators (P2185-02).
The licensee's Design Engineering (DE) group
has completed a design study (ONDS-0180/00, Part A) dated September
30, 1988. The conclusion of this study identified that the increased
weights affected the loads on ten support/restraints (S/R).
The
civil engineering group reviewed this area and determined six S/R
were acceptable and required no changes. One S/R had to be removed to
return the stresses on a pipe to within allowable code requirement Three S/R were required to be modified to bring the stresses within
code requirement The systems associated with the four S/R's -that
required modification were analyzed to be operabl Based on this
action, this item is close.
Exit Interview (30703)
The inspection scope and findings were summarized on January 16,
199 with those persons indicated in paragraph 1 abov The inspectors
described the areas inspected and discussed in detail the inspection
finding The licensee did not identify as proprietary any of the
material provided to or reviewed by the inspectors during this inspection.