IR 05000289/1993013
| ML20045F241 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/30/1993 |
| From: | Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20045F234 | List: |
| References | |
| 50-289-93-13, 50-320-93-06, 50-320-93-6, NUDOCS 9307070125 | |
| Download: ML20045F241 (15) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos.
93-13 93-06 Docket Nos.
50-289 50-320 License Nos.
DPR-50 DPR-73 Lictasee:
GPU Nuclear Corporation
P.O. Box 480
Middletown, PA 17057
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Facility:
Three Mile Island Station, Units 1 and 2
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Location:
Middletown, Pennsylvania Inspection Period:
May 11,1993 - June 21,1993 Inspectors:
Michele G. Evans, Senior Resident Inspector David P. Beaulieu, Resident Inspector Francis I. Young, Senior Resident inspector Approved by:
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Reactor Projects Section [o). 4B, DRP
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John F. Rogge, Chief
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N Insoection Summary: The NRC Staff conducted safety inspections of Unit 1 power operations and Unit 2 cleanup activities. The inspectors reviewed plant operations, i
maintenance, radiological controls, security, and engineering and technical support activities as they related to plant safety.
Results: An overview of inspection results is in ilw executive summary.
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9307070125 930630 PDR ADOCK 05000289
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EXECUTIVE SUMMARY Three Mile Island Nuclear Power Station Report Nos. 50-289/93-13 & 50-320/93-06
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Plant Operations Overall, the inspectors found that shift turnovers were comprehensive and accurate, and adequately reflected plant activities and status. Control room operators effectively monitored plant operating conditions and made necessary adjustments. Housekeeping was satisfactory.
There was extensive management oversight of daily activities. Overall, the licensee conducted Unit 1 plant operations in a safe manner.
The Unit 2 accident generated water (AGW) evaporator continues to operate and approximately 2,004,000 gallons (87% of total) of AGW have been vaporized to the atmosphere at the close of the inspection period.
Due to a plant process computer failure, the licensee missed performing Technical Specification required calculations associated with monitoring of quadrant power tilt and axial power imbalance. The licensee's corrective actions were comprehensive and adequate.
The inspector noted that Plant Operations was operating the emergency diesel generator (EDG) for less than the hour recommended in the operating procedure. Plant Engineering, the Plant Review Group, and the diesel vendor recommended the hour run due to long term reliability concerns. It is a weakness that Plant Operations continues to operate the diesel contrary to the recommended guidance without providing technical justification. The licensee plans to provide technical justification and change the procedure accordingly (URI 50-289/93-13-02).
Due to a personnel error when aligning the 'A' emergency diesel generator for post-maintenance testing, the diesel would not load properly. The licensee plans to enhance the surveillance procedure to better human factor the step that was incorrectly performed.
Radiological Controls During each Auxiliary Building tour, the inspector paid particular attention to ensure
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radiological surveys were current and that proper warning signs were posted. The inspector noted no discrepancies and concluded that overall radiological controls were good.
Due to a design weakness in the locking mechanism on a makeup system filter disposal container, a locked high radiation area was not properly secured. The licensee's evaluation and planned corrective actions are adequate to prevent recurrence.
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Maintenance and Surveillance The inspector found that the licensee's troubleshooting activities involving an electrical failure in the emergency diesel generator protective circuitry were conducted in an excellent manner.
There was good overall coordination between Maintenance, Engineering, and Operations, and good management support for the troubleshooting activities was evident.
Due to a maintenance technician error, a building spray system flow instrument was not properly aligned. As noted in previous NRC Inspection Reports, there have been several recent instances where Instrumentation and Controls technicians have not properly returned equipment to service following maintenance or surveillance activities. The licensee is performing a formal root cause analysis and is preparing a Plant Experience Report to determine if the incidents have a common root cause so that corrective actions can be taken (Unresolved item 50-289/93-13-01).
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TABLE OF CONTENTS Page EXECUTIVE SUM M A R Y....................................... il 1.0 SUMM ARY OF FACILITY ACTIVITIES........................
1.1 Licensee Activities 1-
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1.2 NRC Staff Activities
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2.0 PLANT OPERATIONS (71707, 37828).........................
2.1 Operational Safety Verification
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2.2 Evaporation of Unit 2 Accident Generated Water
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2.3 Plant Process Computer Failure..........................
2.4 Diesel Generator Testing (URI 50-289/93-13-02)
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2.5 Failure to Properly Align Emergency Diesel Generator............
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3.0 RADIOLOGICAL CONTROLS (71707).........................
5-3.1 Failure to Properly Secure locked High Radiation Area....,,.....
'5 4.0 MAINTENANCE AND SURVEILLANCE (61726,62703,71707).........
4.1 Maintenance Observations
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4.1.1 Emergency Diesel Generator Troubleshooting.............
4.1.2 Building Spray Flow Instrument Not Returned to Service (URI 50-289/93-13-01)
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4.2 Surveillance Observations..............................
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5.0 SECURITY (71707)
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6.0 SAFETY ASSESSMENT / QUALITY VERIFICATION (40500)........... 10 7.0 EMERGENCY PREPAREDNESS (71707, 82301)................... 10
8.0 NRC MANAGEMENT MEETINGS AND OTHER ACTIVITIES
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DETAILS 1.0 SUMMARY OF FACILITY. ACTIVITIES i
1.1 Licensee Activities Unit I remained at 100% power throughout the inspection period.
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The Unit 2 Accident Generated Water (AGW) evaporator continued to vaporize AGW to the atmosphere and at the close of the inspection period approximately 2,004,000 gallons (87%
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of total) had been vaporized.
1.2 NRC Staff Activities The inspectors assessed the adequacy of licensee activities for reactor safety, safeguards, and -
radiation protection, by reviewing information on a sampling basis. The inspectors obtained information through actual observation of licensee activities, interviews with licensee personnel, and documentation reviews.
The inspectors observed licensee activities during both normal and backshift hours: 53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> of direct inspection were conducted on backshift. The times of backshift inspection were adjusted weekly to assure randomness.
2.0 PLANT OPERATIONS (71707, 37828)
2.1 Operational Safety Verification The inspectors observed overall plant operation and verified that the licensee operated the plant safely and in accordance with procedures and regulatory requirements. The inspectors conducted regular tours of the follow'mg plant areas:
- Control Room
-- Auxiliary Building
-- Switch Gear Areas
-- Turbine Building
- Access Control Points
-- Intake Structure
-- Protected Area Fence Line
- Intermediate Building i
- Fuel Handling Building
-- Diesel Generator Building The inspectors observed plant conditions through control room tours to verify proper alignment of engineered safety features and compliance with Technical Specifications. The
inspectors reviewed facility records and logs to determine if entries were accurate and identified equipment status or deficiencies. The inspectors conducted detailed walkdowns of accessible areas to inspect major components and systems for leakage, proper alignment, and any general condition that might prevent fulfillment of their safety function.
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The inspector found that shift turnovers were comprehensive and accurate, and adequately reflected plant activities and status. Control room operators effectively monitored plant operating conditions and made necessary adjustments. Housekeeping was commensurate with ongoing work. The inspectors routinely attended licensee management meetings and noted that management was very involved in the daily operation of the plant. The inspector concluded that the licensee conducted overall plant operations in a safe and conservative manner.
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2.2 Evaporation of Unit 2 Accident Generated Water The inspectors observed overall evaporator operation and verified that the evaporator was operated in accordance with licensee procedures and regulatory requirements. At the close of the inspection period, 2,004,000 gallons of the 2.3 million gallons of AGW had been evaporated.
The inspectors identified no conditions that were adverse to safety or contrary to regulatory requirements.
2.3 Plant Process Computer Failure On May 3,1993, while operating at 100% power, the licensee missed the Technical Specification (TS) required calculations associated with monitoring of quadrant power tilt and axial power imbalance. TS 2.5.2.4.g requires the quadrant power tilt to be monitored on a
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minimum frequency of once every two hours when greater than 15% power. TS 3.5.2.8 requires that axial power imbalance be monitored on a minimum frequency of every two hours when greater than 40% power. The plant process computer normally calculates quadrant power tilt and axial power imbalance at six minute intervals. If the plant process computer is unavailable, the licensee performs hand calculations to satisfy the TS
requirements. On May 3,1993, between 1:24 a.m., and 7:30 a.m., the plant process computer failed to initiate the Nuclear Applications Software six-minute calculations.
Therefore, the licensee missed the 3:00 a.m.,5:00 a.m., and 7:00 a.m., hand calculations of quadrant power tilt and axial power imbalance. This incident was properly reported in accordance with 50.73 (a)(2)(i)(b), as an operation or condition prohibited by the plant's TS.
The licensee determined that the plant process computer aborted the six-minute calculation i
each time it was activated and therefore the plant data was not updated and remained
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constant. The shift crew did not realize that the data that they were reviewing was not being updated. At approximately 7:00 a.m., a computer technician noted inoications of a computer malfunction. These indications were not available to control room personnel. At 7:20 a.m.,
the licensee manually transferred to the backup computer. The licensee then performed the required hand calculations based on historical computer data of plant parameters during the
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time of the malfunction. These hand calculations showed that the quadrant power tilt and axial power imbalance remained within the TS limits. The licensee has not been able to determine the cause of the computer malfunction, but believes it may be a hardware problem.
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The only other TS required calculation that was affected by the computer malfunction was the heat balance calculation. This calculation is required to be performed each shift per TS 4.1.
The computer malfunction was discovered in time to perform the heat balance calculation without missing a shift. All the other functions performed by the plant process computer j
were performed properly.
The inspector interviewed the Plant Review Group (PRG) chairman, control room personnel, and the Process Computer Group technician who reviewed this incident. In addition, the inspector reviewed Licensee Event Report (LER)93-004 associated with this incident and found the LER to be comprehensive. The inspector agrees with the licensee that the safety significance of the missed calculations was minimal. Reactor power and imbalance indications are monitored by control room operators and any significant deviation in value that is of sufficient magnitude to challenge TS limits would have been detected. In addition, the follow-up calculations using historical data demonstrated that no TS limits were exceeded.
Therefore, no violation for failing to comply with TS will be issued.
The licensee also evaluated whether the operating crew should have noted the computer malfunction. The data recorded by the control room operator was provided by a computer report that is generated each hour. This hourly report does not provide the date and time of the latest calculation update. The only means that the operator could have noted the computer was malfunctioning was by noting that the effective full power days (EFPD)
number remained the same on the hourly report. The PRG concluded, and the inspector agreed, that it would be unlikely for an operator to notice the unchanging EFPD value.
As a corrective action, the licensee plans to make software changes which detect similar plant process computer malfunctions and alert plant operators that the current calculations are suspect. There will be an alarm generated if a six-minute calculation is older than 12 minutes. In addition, the licensee trained the operating crews on how to recognize when computer calculations may be invalid. The inspector concluded that the licensee's corrective actions were adequate to prevent recurrence of a similar incident.
2.4 Diesel Generator Testing (URI 50-289/93-13-02)
The licensee had taken the 'A' emergency diesel generator (EDG) out of service on June 14,1993, to perform the annual maintenance inspection. Technical Specification (TS) 3.7.2.c requires that the operable EDG be tested daily to very operability when an EDG is out of service. On June 16, 1993, the inspector observed a daily test of the 'B' EDG per Operating Procedure (OP) 1107-03, " Diesel Generator." The inspector noted that although
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OP 1107-03, Step 2.1.1.n, stated that the EDG should be operated at least one hour,' the licensee was operating the EDG for only twenty minutes. OP !107-03 states that operating the EDG for at least one hour allows the engine to reach thermal equilibrium and reduces lube oil and carbon in the exhaust system.
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The inspector discussed the basis for operating the EDG for a period of time less than one hour with Plant Engineering and Plant Operations personnel. The Director of Operations and Maintenance stated that it had been a long term practice to operate the EDGs for twenty minutes to demonstrate operability while the redundant EDG was out of service and that there was a Plant Review Group (PRG) position regarding this.
The PRG Chairman indicated that during PRG Meeting No.91-035 on July 15,1991, the.
PRG held the meeting to help resolve differences between Plant Engineering and Plant Operations on how long the diesel should be operated to verify operability. The issue arose when the Plant Operations was processing Procedure Change Request (PCR) 1-OS-90-058 to OP 1107-03 to clarify the basis for running of the EDG for less than one hour. Plant Operations contended the 20 minute run enhances safety because this minimizes the time the diesel is parallel to offsite power and is unavailable to perform its safety function. However, the P.RG agreed with Plant Engineering and the diesel vendor who indicated that the one hour diesel run was necessary to assure the long term reliability of the diesel by allowing the engine to reach thermal equilibrium and reducing the lube oil and carbon in the exhaust system. The diesel vendor did specify that if a twenty minute run was necessary when the other engine was out of service, then the engine must be run for one hour after the other engine is retumed to service. The Operation's Department did not include the PRG position into the procedure, because they believe it is not always necessary to incorporate vendor guidance. In addition, Plant Operations continued to operate the EDG for twenty minutes, without running the engine for an hour after the other engine was retumed to service. The inspector discussed this issue with Plant Operations management who stated that they plan to reevaluate when the EDG can be operated for less than one hour and will change OP 1107-03 accordingly.
The inspector also discussed TS 3.7.2.c with the licensee which requires the daily diesel
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testing. Generic Letter 84-15, " Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability," encourages licensee's to remove unnecessary diesel testing from TS to improve overall diesel reliability. The licensee plans to review whether removal of TS 3.7.2.c from the TS is appropriate.
The inspector concluded that it was a weakness that Plant Operations continued to operate the diesel in a manner contrary to the guidance provided by the PRG, Plant Engineering, and the vendor without providing adequate technical justification. This issue will remain unresolved pending review of the change to OP 1107-03 (50-289/93-13-02).
2.5 Failure to Properly Align Emergency Diesel Generator On June 18,1993, while performing the post-maintenance testing on the 'A' emergency diesel generator (EDG), the licensee noted that while the diesel was paralleled to off-site power, the diesel electrical load was erratic. Further investigation revealed that the diesel speed droop setting and the unit / parallel switch had not been aligned in accordance with Surveillance Procedure (SP) 1303-4.16, " Emergency Power System."
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The inspector reviewed SP 1303-4.16 and interviewed the Auxiliary Operator who aligned the diesel for operation. The inspector found that the procedure was clearly written, but could be improved to incorporate human factors. SP 1303-4.16, Step 8.1.2, requires the operator to align six different controls. The first four controls are aligned in the control room by a Control Room Opemtor. The last two controls, the speed droop setting and the unit / parallel switch, are to be aligned by the Auxiliary Operator at the EDG. The Auxiliary Operator mistakenly thought that all of the controls were to be aligned from the control room and checked off the step as complete. The inspector discussed the incident with the Director of Plant Operations who' indicated that they plan to incorporate human factors into the procedure l
by splitting Step 8.1.2 into two separate steps. The inspector concluded that this corrective action is adequate to prevent recurrence of a similar incident.
3.0 RADIOLOGICAL CONTROLS (71707)
The inspectors examined work in progress in both units to verify proper implementation of
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health physics (HP) procedures and controls. The inspectors monitored ALARA implemen-I tation, dosimetry and badging, protective clothing use, radiation surveys, radiation protection instrument use, and handling of potentially contaminated equipment and materials. In addition, the inspectors observed personnel working in RWP areas and veri 6ed compliance with RWP requirements. During routine tours of both units, the inspectors veri 6ed a sampling of high radiation area doors to be locked as required. During each Auxiliary Building tour, the inspector paid particular attention to ensure radiological surveys were current and that radiological area postings were consistent with recent surveys. The inspectors noted no discrepancies and concluded that overall radiological controls were good.
3.1 Failure to Properly Secure Locked High Radiation Area On May 16,1993, while performing the weekly veri 6 cation of Locked High Radiation Area integrity, a Radiological Controls technician was able to defeat the locked barrier on the make-up filter disposal container. The barrier had been locked since the last filter disposal on April 7,1993, and the licensee had verified its integrity several times between April 7 and May 15. The licensee found that due to the locking device design and the manner in which
the container cover was locked, it was possible to defeat the locking mechanism and open the container cover.
The licensee reviewed the incident and determined that no violation of a locked high radiation area had occurred. The purpose of the barrier was to prevent inadvertent, unauthorized entry when used with appropriate postings. The barrier was not intended to prevent deliberate, determined attempts to gain unauthorized entry. The licensee concluded that even when locked improperly, the barrier met these criteria. In addition, with the barrier removed the i
maximum contact dose rate was 80 mrem /hr. A jib crane would have been needed to remove the container shield plugs in order to gain access to any field greater than 1000 mrem /hr.
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The licensee's corrective actions included discussion of the incident with appropriate Radiological Control, Operations and Radwaste persormel. The licensee is also pursuing a change to the design 'of the locking mechanism.
The inspector reviewed the licensee's evaluation of the incident and found it to be adequate.
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The inspector interviewed several Radiological Control technicians and found that they were-aware of the locking deFciency and that they knew the correct way to lock the barrier. The inspector concluded that the licensee's actions were appropriate.
4.0 MAINTENANCE AND SURVEILLANCE (61726,62703,71707)
4.1 Maintenance Observations l
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The inspectors reviewed selected maintenance activities to assure that: the activity did not violate Technical Specification Limiting Conditions for Operation and that redundant components were operable; required approvals and releases had been obtained prior to
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commencing work; procedures used for the task were adequate and work was within the skills
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of the trade; maintenance technicians were properly qualified; radiological and fire prevention i
controls were adequate; and, equipment was properly tested and returned to service.
Maintenance activities reviewed included:
Job Order Number 074011, " Replace LS-24 Cam on West Side of Spent Fuel Pool,"
on June 8,1993.
Job Order Number 074010, " Remove and Inspect the Southwest Bridge Wheel," on
June 8,1993.
j Job Order Number 073525, " Control Rod Mast Electrical Cable Reel Spring Tension is
Weak," on June 8,1993.
Job Order Number 070065, " Electrical Maintenance Emergency Diesel Generator
Work," on June 14, 1993.
Job Order Number 070077, " Mechanical Maintenance Emergency Diesel Generator
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Work," on June 14, 1993.
Refueling Procedure 1503-1, " Receipt of New Fuel and Control Components," on
May 17,1993.
Corrective Maintenance Procedure 1420-Y-13, " General Circuit Troubleshooting and
Repair," on June 10, 199.
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Overall, the inspectors found that individuals involved in maintenance activities were knowledgeable and work was conducted using appropriate procedures. The inspectors found that the receipt inspection and testing of new control rod assemblies was thorough, with good supervisory oversight. The inspectors also observed good supervisory oversight during the EDG annual inspection. Review of the licensee's activities associated with the troubleshooting of an electrical failure with the 'B' EDG is described further in Section 4.1.1.
4.1.1 Emergency Diesel Generator Troubleshooting On June 10, 1993, the inspector observed the licensee activities associated with an electrical failure in protective circuitry that automatically shuts down the emergency diesel generator (EDG) when an undesirable condition is sensed. The licensee became aware of the electrical failure when an Auxiliary Operator noted that an engineered safeguard status light in the 'B'
EDG protective circuitry was out. A thyristor in the circuitry for lubricating oil low temperature shorted, causing a fuse in the EDG shutdown protective circuitry to blow. The blown fuse disabled other EDG protective features including, low lubricating oil pressure protection, high crankcase pressure protection, engine overspeed protection, and various other alarm features. Plant maintenance attempted to replace the blown fuse, but the new fuse also blew. Plant Maintenance conducted troubleshooting in accordance with Corrective Maintenance Procedure 1420-Y-13, " General Circuit Troubleshooting and Repair."
Shortly after this condition emerged, the Manager of Plant Electrical Engineering and the Director of Plant Operations were involved in assessing the condition of the diesel. The licensee determined that even though the EDG would be able to start and load properly, it should be declared inoperable due to non-functional diesel protective features. The 'B' EDG was declared inoperable and the 'A' EDG was started to verify operability in accordance with the TS. Since a replacement thyristor was unavailable, Plant Engineering wrote an Engineering Evaluation Report to allow the use of an equivalent component. The thyristor was replaced and the diesel was returned to service later that day.
The inspector discussed the status of the EDG with the Shift Supervisor and independently verified that there was no inoperable engineered safeguards equipment in the 'A' train which would require the licensee to enter TS 3.0.1 (General Action Requirement.) Plant Operations reviewed the previous Auxiliary Operator's logs and verified that the engineered safeguards status light was on during the previous shift.
The inspector concluded that conduct of the EDG troubleshooting activities was excellent.
There was good overall coordination between, Maintenance, Engineering, and Operations and good management support for the troubleshooting activities was evident.
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4.1.2 Building Spray Flow Instrument Not Returned to Service (URI 50-289/93-13-01)
On June 8,1993, while performing Surveillance Procedure 1300-3A, " Inservice Test of Building Spray (BS) Pump 1 A/B and Valves," the flow for BS pump 'lB' indicated lower than expected. BS flow instrument, BS-1-DPT-2, saw 153 psid and the required action range for this instrument is 172 psid. The licensee found that equalizing valve, BS-V-1023, on the BS How instrument was open. The last performance of SP 1300-3A was completed satisfactorily on March 8,1993. BS-1-DPT-2 was subsequently calibrated on March 17,1993. The licensee believes that the equalizing valve was left open by an Instrumentation & Controls (I&C) technician at that time. When the licensee closed the equalizing valve, instrument differential pressure returned to the expected value.
There is no TS which applies specifically to the BS flow instrument. Following a postulated large break loss of coolant accident (LOCA), the BS pumps and low pressure injection (LPI)
pumps take suction from the borated water storage tank (BWST) to perform their safety function. When the BWST level drops to the lo-lo level alarm setpoint (6 feet 4 inches),
operators align the suction for the BS and LPI pumps to the reactor building sump. The BS and LPI pumps for each train share a common suction line from the reactor building sump.
The PRG performed an operability determination on the 'B' BS train with the flow instrument inoperable. The only function of the BS flow instrument is for control room indication.
Abnormal Transient Procedure 1210-7, "Large Break Imss of Coolant Accident Cooldown,"
requires that operators throttle the BS pumps to 1300 -1400 gpm when on recirculation from the reactor building sump. FSAR Section 6.4.2, indicates that throttling BS flow is necessary to ensure adequate net positive suction head (NPSH) for the BS and LPI pumps (common suction line) when the reactor building sump water is in a saturated condition (i.e. not taking into account reactor building overpressure). With the BS flow instrument falsely indicating lower than actual flow, operators would not throttle the 'B' BS train flow to 1400 gpm.
Therefore, potential NPSH problems would exist for the BS and LPI pumps. When initiating sump recirculation flow, the maximum BS flow with no throttling would be 1540 gpm. The licensee performed NPSH. calculations which took in account the reactor building pressure that would exist following a LOCA. The licensee calculated that, for the worst case conditions of reactor building pressure and sump water temperature, the actual NPSH available to the BS and LPI pumps would be approximately 26 to 27 feet of water. The amount of NPSH necessary for reliable pump operation is 14 to 15 feet. When Reactor Building pressure decreases to 4 psig, operators secure the BS pumps, which eliminates the NPSH concern.
Based on this evaluation, the PRG concluded that the BS and LPI systems would have been able to perform their specified functions if called upon to do so and that the systems remained operable during the time the equalizing valve was open. Therefore, the PRG concluded that this event was not reportable. Since the BS and LPI systems would not meet the FSAR design conditions concerning the ability to operate BS and LPI pumps with a saturated reactor building sump, this event resulted in a degraded or nonconforming condition.
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The i'nspector interviewed the I&C technician who aligned the BS flow detector following calibration. The I&C technician calibrated the BS flow instrument in accordance with Preventive Maintenance Procedure (PMP) IC-1, " Flow Loop Calibration." Step 8.7.3 clearly states to close the equalizing valve. The equalizing valve is physically located behind the manifold. The I&C technician indicated that due to the valve orientation, he may have inadvertently turned the hand wheel in the wrong direction. PMP IC-l' does not require an independent verification of valve position.
There have been several recent instances where I&C technicians have not properly returned'
equipment to service following maintenance or surveillance. This equipment includes the reactor building atmospheric monitor, once-through-steam-generator level instruments',
instrument air to the pressurizer level control valve, and instrument air valves to steam driven emergency feedwater pump steam pressure controller. The licensee is performing a formal root cause analysis and is preparing a Plant Experience Report to determine if the incidents have a common root cause so that corrective actions can be taken.
The inspector reviewed the licensee's operability and reportability determinations and found them to be acceptable. The recent instances of I&C technicians not properly returning equipment to service is a safety concern and may be indicative of a programmatic weakness.
This issue will remain unresolved pending inspector review of the licensee's Plant Experience Report (50-289/93-13-01).
4.2 Surveillance Observations The inspe: tors observed conduct of surveillance tests to verify that approved procedures were being used, test instrumentation was calibrated, qualified personnel were performing the tests, and test acceptance criteria were met. The inspectors verified that the surveillance tests had been properly scheduled and approved by shift supervision prior to performance, control room operators were knowledgeable about testing in progress, and redundant systems or components were available for service as required. The inspectors routinely verified adequate performance of daily surveillance tests including instrument channel checks and reactor coolant system leakage measurement.
Surveillance activities reviewed included:
Surveillance Procedure ll.39A, " Heat Sink. Protection System - EFW Automatic
Initiation," on June 4,1993.
Surveillance Procedure 1302-3.1, " Radiation Monitoring System Calibration," on e
June 8,1993.
Surveillance Procedure 1300-3A, " Inservice Test of Building Spray (BS) Pump 1 A/B
and Valves," on June 15, 1993.
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Surveillance Procedure (SP) 1303-4.16, " Emergency Power System," on
June 18,1993.
Operating Procedure 1107-3, " Diesel Generator," on June 16, 1993.
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Overall, the inspectors found that surveillance activities were performed in a controlled
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manner using appropriate procedures. During the performance of SP 1302-3.1, the I&C technicians kept the control room operators well informed when an alarm was expected as a result of testing. During the previous performance of SP ll.39A on March 12,1993, a reactor trip had occurred due to an electrical failure in the reactor coolant pump power monitor. During the June 4,1993 performance, all equipment functioned as designed and the surveillance was accomplished without incident. Inspector review of EDG testing is discussed further in Section 2.4.
5.0 SECURITY (71707)
The inspectors monitored security activities for compliance with the accepted Security Plan and associated implementing procedures. The inspectors observed security staffing, operation of the Central and Secondary Alarm Stations, and licensee checks of vehicles, detection and assessment aids, and vital area access to verify proper control. On each shift, the inspectors observed protected area access control and badging procedures. In addition, the inspectors routinely inspected protected and vital area barriers, compensatory measures, and escon procedures. On June 10, 1993, the inspector noted that appropriate security measures were taken when a repair truck was parked next to the protected area fence. The inspectors concluded that overall the Security Plan was being properly implemented.
6.0 SAFETY ASSESSMENT / QUALITY VERIFICATION (40500)
On June 9,1993, the inspectors attended a General Office Review Board (GORB) meeting.
The inspectors observed discussions concerning recent Unit 1 operating activities and Unit 2 Post-Defueling-Monitored-Storage preparation activities. The inspector found that the GORB's review of the issues discussed was comprehensive and there was a focus on safety.
7.0 EMERGENCY PREPAREDNESS (71707,82301)
On May 19,1993, the licensee conducted its annual emergency preparedness exercise. The results of the inspection will be discussed in Inspection Report 50-289/93-10.
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8.0 NRC MANAGEMENT MEETINGS AND OTIIER ACTIVITIES At periodic intervals during this inspection, meetings were held with senior plant management to discuss licensee activities and areas of concern to the inspectors. At the conclusion of the reporting period, the resident inspector staff conducted an exit meeting with licensee management summarizing inspection activities and findings for this report period. Licensee comments concerning the issues in this report were documented in the applicable report section. No proprietary information was identified as being included in the report.
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