IR 05000219/1990007
| ML20043G145 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/25/1990 |
| From: | Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20043G144 | List: |
| References | |
| 50-219-90-07, 50-219-90-7, NUDOCS 9006190117 | |
| Download: ML20043G145 (69) | |
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O U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-219/90-07 Docket No.
50-219 License No.
OPR-16 Priority --
Category C Licensee:
GpU Nuclear Corporation i Upper Pond Road Parsippany, New Jersey 07054 Facility Name: Oyster Creek Nuclear Generating Station Inspection Conducted: March 18, 1990, - April 21, 1990 Participating Inspectors:
M. Banerjee, Resident Inspector E. Collins, Senior Resident Inspector D. Lew, Resident Inspector N. McNamara, Laboratory Ass't, DRSS R. Struckmeyer, Sr, Rad. Spec., DRSS r
Approved By:
rM M
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W. Ruland, Actinc Section Chief,
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, Date Reactor Projects Section 4B l
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Inspection Summary:
I Inspection Report No. 50-219/90-07 for March 18, 1990 - April 21, 1990 Areas Inspected:
The inspection consisted of 203 hours0.00235 days <br />0.0564 hours <br />3.356481e-4 weeks <br />7.72415e-5 months <br /> of direct inspection by resident inspectors. The areas inspected included observation and review of plant operational events (paragraph 1.0); review of radiological events (paragraph 2.0); routine observations of maintenance activities and surveillance tests (paragraph 3.0); update of drywell corrosion (paragraph 4.1); engineering evaluations of core spray valve operating times (paragraph 4.2); observation of an emergency drill (paragraph 5.0); routine security
observations (paragraph 6.0); review of corrective actions associated with vendor quality (paragraph 7.1); review of safety evaluation for multiple
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control rod motion (paragraph 7.2); and, Temporary Instruction 2500/27 i
(paragraph 7.3).
Results: Overall, the plant was operated in a safe manner.
Two unresolved
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items were opened, and one unresolved item was closed. An executive summary i
follows, 9006190117 900607 j
PDR ADOCK 05000219 l
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TABLE OF CONTENTS P_ag I.
Executi ve Summa ry........................ iii II.
Details.............................
1.0 Operations...........................
I 1.1 Chronology of Operational Events (71707, 93702)......
I 1.2 Injection of Halon Suppression System in the Site Emergency Building (71707, 93702)...........
1.3 Control Room Tours (71707).................
1.4 Facility Tours (71707)...................
i 2.0 Radiological Controls......................
2.1 New Radwaste Heat Exchanger (71707), 93702)........
2.2 Auxiliary Boiler system Deaerating Feed Tank Spill (93702).
2.3 Radiological Planning for Reactor Recirculation Pump Seal (71707).....................
2.4 Ingestion of a Hot Pa,rticle (93702)............
3.0 Maintenance / Surveillance....................
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3.1 Monthly Maintenance Observation (62703)..........
3.2 Monthly Surveillance Observation (61726)..........
4.0 Engineering and Technical Support.
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4.1 Drywell Corrosion (71707, 30702)..............
10-4.2 Core Spray Valve Stroke Time (93702)...........,
5.0 Emergency Preparedness (71707).................
- 6.0.0bservation of Physical Security (71707)............- 11 7.0 Safety Assessment / Quality Verification (92702).........
7.1 Main Steam Isolation Valve NS03A Leak Repair........
7.2 Safety Evaluation for two Control Rods Moving at the Same Time (71707), 92700).................
7.3 Temporary Instruction 2500/27 (25027)...........
7.4 Management Meeting (30702).................
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P_agg 8.0 Inspection Hour Summary.....................
9.0' Exit Meeting and Unresolved Items (30703)............
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9.1 Preliminary Inspection Findings..............
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9.2 Unresolved Items.......................
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I ATTACHMENTS
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List of Personnel Contacted Attachment II: Management Meeting, April 6, 1990-O,
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
EXECVIIVE SUMMARY Report No.
50-219/90-07 plant Operations
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Overall, the plant was operated in a safe manner.
During power ascension the-
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operators selected two control rods simultaneously and responded by deenergiz-
ing rod select power, which deselected the rods.
No rod motion occurred.
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Compensatory actions were implemented to ensure that only one control rod at a time is selected and moved. The licensee commenced a reactor shutdown on March 26 and began an eight day outage for repair of main condensed air ejection valves, a reactor recirculation pump seal, and to obtain drywell shell thick-ness measurements.
One of the two emergency diesel generators was declared
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inoperable for a four-day period due to battery problems. At the end of the inspection period, the reactor was shutdown and the plant was entering an un-planned outage due to loss of one of the two safety related unit substations.
Licensee initial actions in response to the substation loss were appropriate.
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Radiological Protection Inspection of the auxiliary boiler contamination was completed. Minor
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contamination of normally clean areas was found and decontaminated. An
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ingestion of a hot particle resulted primarily from poor radiological i
practices. Work planning and radiological controls were improved during the reactor recirculation pump seal replacement relative to a pr.ior replacement in
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February.
Surveillance and Maintenance Both surveDiance testing and maintenance activities were observed.
No deficiencies were identified.
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Emergency Preparedness A dress rehearsal drill for the annual exercise was observed.
No unacceptable-i conditions were identified.
Security Routine inspections did not identify any deficiency.
Engineering and Technical-Support Licensee representatives discussed the results of the drywell thickness measure-ments taken-during the 12-U-J outage with the NRC.
The licensee's previous-conclusion that the drywell structural integrity is assured until August 1991 has not changed.
Safety Assessment / Quality Verification Licensee review and compensatory measures addressing a vulnerability of the-reactor manual control system were appropriate and provide additional assurance that multiple control rod motion will be prevented.
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DETAILS 1.0 Plant Operational Review 1.1 Chronology of Operational Events The inspectors reviewed det6ils associated with key operational events that occurred during the report period. A summary of these inspection activities follows.
At the beginning of the inspection period the reactor power was 58%,
limited by condenser vacuum problems.
The licensee continued to investi-gate to determine the root cause of the problems and neca'::..y corrective action.
3/22/90 The licensee entered a seven-day technical specification
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action statement when "B" control rod drive (CRD) pump was tagged out of service for an oil leak repair. This repair was completed the
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same day and the "B" CRD pump was declared operable.
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3/23/90 The #2 dilution pump became unavailable after a light socket
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in the automatic bypass switch shorted during a light bulb change.
The short was repaired and the pump was returned to service the same day.
The "A" CRD pump was removed from service for maintenance ano the licensee entered a seven-day technical specification LCO.
The maintenance was completed, and the CRD pump was declared operable the same day.
3/26/90 At 10:00 p.m. the licensee commenced a controlled reactor
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shutdown to repair steam jet air ejector problems which were causing low condenser vacuum and to investigate the "A" recirculation pump seal which had indications of a possible failure.
During the shutdown IRM 12 was bypassed due to spiking and declared inoperable.
Also, valve packing adjustments were performed on isolation condenser steam valve V-14-32.
3/27/90 Cold shutdown was reached at 11:50 a.m.
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3/29/90 While performing a survey inside the drywell, a radiological
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control technician ingested a hot particle. This event is discussed in paragraph 2.4.
3/30/90 IRM 12 was declared operable.
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4/3/90 At 10:35 p.m. the reactor mode switch was placed to startup
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and the licensee commenced pulling the control rods.
Criticality was achieved the same day at 11:40 p.m.
Startup was observed by a
resident inspector.
No significant problems were experienced by the licensee.
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During the heatup, IRM 11 failed downscale and was bypassed.
4/4/90 A drywell leak inspection at 1000 psi reactor pressure
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i was completed and main steam isolation valve (MSIV) NS03A was injected with Leak Repair to minimize the flange leak.
At 3:54 p.m. with the reactor at 90% power, immediately after a power surge in the grid, the fire alarm and halon discharge actuated in the site emergency building (SEB) computer room.
Prime plant computer, rod worth minimizer (RWM), and safety parameter display system (SPDS)
were lost es a result of this. The licensee made the necessary noti-fication and established an hourly fire watch. Within two hours the computers, including RWM and SPDS, were returned to normal and the licensee commenced increasing power.
Details of this incident are discussed in paragraph 1.2.
During reactor power ascension the operators selected two rods simul-taneoutiy and responded by deenergizing rod select power, which de-selected the rods.
No rod motion occurred. The licensee submitted a deviatich report. The capability to select more than one rod at the i
same time has been previously identified and reviewed in Inspection i
Report 50-219/89-29.
The safety evaluation is discussed in paragraph
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4/5/90 The reactor mode switch was placed in RUN at 12:12 a.m., and
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the main generator was placed on line the same day.
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During control rod scram testing, rod 34-07 indicated " overtravel in" and could not be returned to the "00" position.
The control rod
was declared inoperable and was left fully inserted.
4/7/90 The plant reached 100% power.
During the power ascension the i
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licensee experienced average power range monitor (APRM) rod blocks in
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channel 1 at a level lower than the 104% setpoint.
The cause was
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identified to be a defective flow converter in APRM system 1.
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4/9/90 Reactor power was reduced to 80% and the APRM system 1 flow
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converter was replaced.
The power was increased to 100% the same day.
The Standby Gas Treatment System (SGTS) II was removed from service for sample canister replacement. A seven-day technical specificction action statement was entered.
The system was returned to service the same day and the action statement was terminated.
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4/12/90 The Standby Liquid Control System (SLCS) was taken out of
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service to replace a defective photoelectric cell in the system I
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continuity meter which was causing false alarms. The plant was in a seven-day technical specification action statement. The SLCS was declared operable on 4/13/90, following the repair of the continuity
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meter.
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4/15/90 The acoustic flow monitor NR28P was declared inoperable
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during a surveillance test. As required by the plant technical spe-cifications, the gain on the adjacent acoustic monitor NR26N was increased.
Following a review of the test data by engineering, the monitor was declared operable on 4/16/90 and the gain on the adjacent monitor was returned to its normal value.
4/16/90 During a surveillance the emergency diesel generator (EDG)
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failed to start. The EDG was declared inoperable and a technical specification action statement was entered.
After troubleshooting and repair the EDG was returned to service on 4/20/90.
Inspector review of EDG battery problems was ongoing at the end of the inspection period.
4/21/90 At 9:54 a.m., while at full power, the plant experienced a
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loss of unit substation USS 182, one of two 460V safety related substations. As a result the licensee commenced a controlled reactor shutdown at 10:50 a.m.
Also, as a result of this event, the licensee was unable to pump down the dr/well equipment drain tank which overflowed to the drywell sump.
ihe indicated drywell unidentified leakage increased above 5 gpm and the licensee declared an unusual event.
Inspector review of this event was ongoing at the end of this inspection period.
1.2 Injection of Halon Suppression System in the Site Emergency Building On April 4,1990, immediately following a power surge, the fire alarm and the halon system actuated in the site emergency building (SEB) computer room. Due to this halon injection the plant computer, including both the Safety Parameter Display System (SPDS) and Rod Worth Minimizer (RWM), were lost. At the time of the incident the plant was in startup at approxi-mately 4% power.
The licensee halted the plant startup and assessed the incident.
The power increase was restarted once the plant computer, SPDS, and RWM were returned to service. An immediate notification was made to the NRC due to loss of emergency assessment capability.
The licensee investigated the cause of halon injections and determined it to be personnel error.
Following the power surge, an SEB trouble alarm was received in the fire panel at the guard house. A security guard responding to the alarm inadvertently activated the halon injection. The security guards were previously instructed to call the co'ntrol room upon an SEB trouble alarm.
The licensee has performed a security investigation and an HPES evaluation and is currently evaluating any needed corrective actions. The inspector concluded this was an isolated event and found the licensee's followup action adequate.
1.3 Control Room Tours Routine tours of the control room were conducted by the inspectors during which time the following documents were reviewed:
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Control Room and Group Shift Supervisor's Logs;
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Technical Specification Log;
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Control Room and Shift Supervisor's Turnover Check Lists;
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Reactor Building and Turbine Building Tour Sheets; l
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Equipment Control Logs;
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Standing Orders; and,
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Operational Memos and Directives.
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No unacceptable conditions were identified, i
1.4 Facility Tours Routine tours of the facility were conducted by the inspectors to make an
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assessment of the equipment conditions, personnel safety, and procedural
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adherence and regulatory requirements. The following areas were among l
those inspected:
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Turbine Building
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Vital Switchgear Rooms j
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Cable Spreading Room
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Diesel Generator Building
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Reactor Building l
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New Radwaste Building l
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Old Radwaste Building
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Fire Protection:
Randomly selected fire extinguishers were accessible and
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inspected on schedule.
l Fire doors were unobstructed and in their proper position.
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Ignition sources and combustible materials were controlled in
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accordance with the licensee's approved procedures.
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Equipment Control:
Jumper and equipment mark-ups did not conflict with technical
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specification requirements.
Conditions requiring the use of jumpers received the prompt
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attention of the licensee, l
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Vital Instrumentation:
Selected instruments appeared functional and demonstrated
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parameters within Technical Specification Limiting Conditions for Operation.
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Housekeeping:
Minor housekeeping deficiencies which were identified were promptly
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corrected by the licensee. No other unacceptable conditions were identified.
2.0 Radiological Controls 2.1 New Radwaste Heat Exchanger The inspector reviewed a completed work package (Job Order No. 19507) on the New Radwaste Heat Exchanger.
The work involved preventative maintenance on the service water side of the heat exchanger.
In the work package, a maintenance worker had annotated that a step was not performed because his supervisor ordered him to close the heat ex-changer without performing the step. The step not performed was a swipe of tne heat exchanger prior to closing it up.
This action was contrary to the procedural and radiological controls requirements. The inspector con-ducted further review of this event based on a potential-concern that pro-cedural requirements were not being followed and that the apparent attitude
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toward procedural requirements may be poor.
The inspectors interviewed the supervisor and worker involved in this event.
Based upon these interviews the following conclusions were reached:
The supervisor did direct the maintenance worker to close the heat
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exchanger, The supervisor did not recognize at the time that he was violating
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the work procedure; and,
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Later that shift the supervisor recognized the potential for an
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error. The next shift opened the heat exchanger and took the swipe.
The inspectors concluded that a contributing cause for this event was miscommunication between the maintenance worker and the supervisor. The miscommunication was a result of a poor relationship between the indivi-duals.
No evidence of a Mor attitude toward procedural requirements was found in this event. The mistake which occurred in this event was prompt-ly corrected. No unacceptable conditions were identified.
2.2 Auxiliary Boiler System Deaerating Feed Tank Spill The licensee performed an evaluation of the potential offsite concentra-tions of radionuclides that might have been released from the deaerating feed tank spill of March 16 (50-219/90-06).
The licensee concluded that these concentrations would represent a very small fraction of the limits in 10 CFR 20 Appendix B.
The inspector reviewed these calculations and determined the licensee's conclusions to be acceptable.
Inspection Report 50-219/90-06 opened an unresolved item regarding review
of the licensee's 50.59 evaluation for continued operation of the auxil-iary boiler system with contaminated water (50-239/90-06-04).
Regional and resident inspectors re'viewee revision 0 and revision 1 of the safety evaluation.
This evaluation estimated the total possible activity content of the system and identified potential offsite release paths. The off-
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site consequences of tube failures and liquid releases were evaluated.
For the worst case tube rupture, the licensee concluded the integrated offsite dose was 2.27 mrem.
For the worst case liquid release, the li-censee concluded that the concentr0tions of radioactive material in the canal at the Route 9 bridge would be 54% of the 10 CFR 20, Appendix B, Table II limits.
The above calculations assumed an activity in the auxil-iary boiler system of 1.65 E-2 microcuries per milliliter.
No unaccept-able conditions were identified. Unresolved item 90-06-04 is closed.
To discuss the auxiliary boiler contaminations and continued operation of the auxiliary boiler system with contaminated water, a conference call was conducted on March 21, 1990, between the Oyster Creek Site Director and NRC Region I management.
Item discussed included the following:
Actions taken to prevent further releases;
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Completion of the safety evaluation;
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Decontamination of the auxiliary boilers;
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Identification of other potential release paths, and the monitoring
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requirements necessary to prevent further releases; l
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Consequences of a boiler tube rupture; and,
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Compliance with the site technical specifications for liquid effluent
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discharges.
No unacceptable conditions were identified.
Inspectors reviewed the licensee's response and corrective actions associated with radioactive contamination found on the turbine building roof.
As a result of followup actions from the contamination incident on March 22, 1990, the licensee identified small levels of loose surface contamination on the northwest turbine building roof. Additional surveys identified additional low levels of contamination on the southwest turbine building roof.
Both areas contained steam heating coils which are sup-plied from the site auxiliary boiler system.
Low levels of contamination were also identified on leaky valves associated with the heating coils.
Additional heating coils in the main office building and new radwaste building were surveyed and found not to be contaminated.
Previously, the licensee had identified the steam heating coils on building roofs as potential release paths for contamination from the auxiliary boiler system. The licensee had sampled condensate from the heating coils for activity. These results were negative.
The steam heating system has not been in service since the auxiliary boiler contaminations.
Licensee surveys of the northwest turbine building roof drain pipe identified contamination levels of approximately 1,500 counts per minute.
The portion of the drain pipe which could be removed was removed, and other portions of the drain system which were accessible were decontaminated.
No activity was identified in the drains associated with l
the southwest turbine building roof.
Licensee evaluation concluded that l
because of the low levels of contamination any potential release outside the radiological controls area was incalculably small and would have been
indistinguishable from background.
The licensee also concluded there were no off-site consequences as a result of this contamination.
At the end of the inspection period the licensee had blocked the storm drain associated with the northwest turbine building roof to prevent the l
release of contamination off site.
Decontamination efforts were being implemented on both the northwest and southwest turbine building roofs.
Additional surveys in the ventilation systems did not identify any contamination.
The source of the contamination was identified by the licensee as the site auxiliary boiler system. During the period of operation of the system with high levels of radioactive material, leaking steam isolation valves I
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allowed low levels of contamination to accumulate over time at the heating coils. Once contamination levels in the boiler system were reduced, activity spread to the heating coils was minimized.
Leakage from valves is being collected to preclude further spread of contamination.
Inspectors had no further questions.
2.3 Radiological Planning for Reactor Recirculation Pump Seal Replacement Inspectors reviewed the radiological planning and actual radiation exposures associated with the replacement of the "A" reactor recirculation pump seal during outage 12-U-J, which lasted from March 26 to April 3.
The licensee estimated the radiation exposure associated with the seal replacement would be 5 man-Rem and that the exposure which would be incur-
red because of bearing work would be 4 man-Rem. These estimates were based on historical experiences for this work.
Because of the high ex-posure associated with this job when it was previously performed (Inspection Report 50-219/90-06) the licensee decided to install more shielding. The radiation exposure incurred as a result of shielding and scaffolding installation was less than 2 man-Rem.
The actual radiation exposure incurred as a result of seal replacement and bearing work was 4.19 man-Rem.
For the seal, bearing, shielding, and scaffolding work the total exposu're was approximately 6 man-Rem, about 3 man-Rem less than es-timated. The inspector concluded this reduction in exposure was due to better work planning and better coordination between the radiation pro-tection staff and the work organization. Additional exposure of approxi-mately 3 man-Rem, for a total of 9 manRem, was incurred because of addi-tional work scope, including metal chips inspection, flushing the seal cooler, recirculation impeller inspection, and post maintenance testing.
No unacceptable conditions were identified.
2.4 In,gestion of a Hot Particle During a radiological survey inside the drywell on March 29, 'i90, two radiological controls technicians experienced facial contamination, and one technician ingested a hot particle.
The technicians had been performing radiological surveys in the upper portion of the drywell and the reactor pressure vessel head region. The reactor vessel head area is accessed by a manway. One technician, feeling faint from the heat in the area, exited the manway. The second technician performed the survey, and while he was exiting the manway he was assisted from below by the other technician.
It was this locating of one person above the other in a high contamination area which allowed facial contamination and subsequent ingestion of the hot particle.
The hot particle was detected during subsequent personnel surveys and a whole body count was performed.
Because of the movement of the particle from the upper chest to the abdomen region, the licensee concluded the particle had been ingested.
The particle was subsequently collected for evaluatio *
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The particle was measured and determined to have an activity of 2.3 microcuries.
The licensee performed dose estimates and assigned approximately 30 maximum permissible concentration (MPC) hours and 216 mrem dose to the intestinal tract.
Prior to performing the survey in the drywell the use and need of respiratory protection was evaluated by the licensee but was ruled out because of the opposing concerns associated with heat and stress in the upper portions of the drywell. A dust mask was not used because one could not be located.
The cause of the ingestion of the hot particle was the poor practice in a high contamination area of one worker being located physically below another worker. The movement and dislodging of radioactive material caused one technician to ingest the hot particle.
In addition, the use of a dust mask may have prevented this ingestion.
Licensee corrective actions in response to this event were still being formulated at the end of the inspection period.
The inspector had no other questions.
3.0 Maintenance / Surveillance
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3,1 Monthly Maintenance Observations
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On March 23, 1990, the inspector observed partial ccmpletion of a pressure switch installation on fire diesel #2. Oyster Creek has two fire diesels which run the two main fire pumps.
1he pumps actuate upon a low fire header pressure.
The licensee implemented a modification to install pressure switches of different design after the original one was broken on fire pump #2 and could not be returned to service.
The inspector reviewed the york _ptekage and installation procedure. The work was properly classifieo, QC hold points instituted, and replacement components had procurement QA tags.
The inspector did not identify any unacceptable conditions.
On April 20, 1990, maintenance work on the #1 emergency diesel generator was observed. The maintenance involved replacement of degraded battery cells identified during a diesel surveillance.
The inspector reviewed the work package for the battery cell replacement.
Appropriate use and adherence to the work package was observed.
3.2 Monthly Surveillance Observation Inspectors observed execution of the following surveillances:
619.3.001 - Turbine Load Rejection Scram Test 665.5.005 - Drywell Air Lock Leak Rate Test
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The inspector verified that appropriate aporoval was obtained and prerequisites were verified prior to starting the procedures.
Test equipment used was as specified and calibrt.ted.
During performance of the first procedure, chattering of the relays associated with 40% power scram bypass logic was observed. This phenomenon has been attributed to the vibration of the Barksdale pressure switches (PSH) at the turbine front standard.
The licensee concluded that this relay chattering did not affect scram setpoint or bypass setpoint.
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The inspector verified that a valve verification checkoff was performed by appropriate personnel who did not perform or witness previous valve lineups or manipulations during the test.
Followup actions to address deficiencies were appropriate.
During performance of the second surveillance procedure, the technicians did not sign-off one work step at the time of completion and did not enter final data on the control copy of the procedure at the time of collection.
Instead, data was written on a small piece of note paper. The data were conveyed to the involved supervisor for leak rate calculation.
Afterwards, the step was signed off and the data entered into the procedure test copy.
This practice deviates from the licensee's established guideline that necessary data should be recorded as the task is performed.
Inspectors discussed this methodology with the technician and his supervisor.
During this test, data is observed until conditions stabilize.
The last data point, after stabilization, is the test data.
The technician acknowledged it should have been recorded at that point in time.
Inspectors concluded the correct value was recorded into the procedure and that these weaknesses did not adversely affect the conclusions of the test.
Inspectors had no other questions.
4.0 Engineering and Technical Support 4.1 Drywell Corrosion The licensee performed drywell thickness measurements during the eight-day (March 26 to April 3) outage.
This included the following:
Three locations at elevation 81 ft, which were not inspected in
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February 1990; An A-scan of the accessible portions of the drywell
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circumference at elevation 50 ft., including UT readings at the three thinnest locations found by the A-scan; Areas of the sand bed not inspected since October 1988 or areas
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that showed changes in corrosion rate in February 1990 data.
In a telephone conference on April 4, 1990, the licensee provided the drywell inspection results to the NR *
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The licensee's evaluation of the data indicated the overall conclusions of the previous safety evaluation were unchanged.
Based on an increased corrosion rate at the 50 ft, elevation identified during the last two inspections, the licensee has projected the drywell structural integrity is assured until August 1991. A letter from GPU to the NRC dated April 11, 1990, reported the inspection results and the licensee's plans regarding activities to abate drywell corrosion, analyze the drywell, and develop methods for needed drywell repair.
i 4.2 Core Spray Valve Stroke Time During January 1990, licensee review of the valve IST program against Generic Letter 89-04 identified a potential concern with the stroke times of some core spray system valves.
The core spray test return valves were
found to have a stroke time which exceed Appendix K assumptions for i
establishing full core spray flow. The test return valve is normally j
closed and is open only during system testing. Thus if a loss of coolant (LOCA) occurred during testing, this valve would be required to close to prevent core spray flow diversion. The licensee wrote a deviatiun report and was evaluating the impact on Appendix K analysis. This item will be i
unresolved pending completion of the licensee's evaluation.
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(50-219/90-07-01)
5.0 Emergency Preparedness i
On March 21, 1990, the licensee conducted an emergency preparedness drill.
This drill was a dress rehearsal for the annual exercise.
The inspector i
reviewed the drill scenario and observed the drill from the control room and the Technical Support Center.
No unacceptable conditions were identified.
6.0 Observation of Physical Security During daily tours, the inspectors verified that access controls were in l
accordance with the Security Plan, security posts were properly manned, protected area gates were locked or guarded and that isolation zones were free of obstructions. The inspectors examined vital area access points to
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verify that they were properly locked or guarded, and that access control i
was in accordance with the security plan.
7.0 Safety Assessment / Quality Verification
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7.1 Main Steam Isolation Valve NS03A Leak Repair The main steam isolation valve NS03A inside the drywell had a body to
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bonnet leak that was repaired with a clamp and Fermanite 2X sealant by Leak Repair, Inc. during the February outage.
It was later identified that the licensee had not performed a receipt inspection on the Fermanite 2X
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material and the clamp.
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The licensee's corrective action for missed receipt inspection on Fermanite was reviewed by the inspector in Inspection Report 50-219/90-06.
This consisted of doing a laboratory analysis of material obtained from the vendor with the same batch number and procuring a refrigerator to store a supply of this material on site.
During the eight-day outage another application of the sealant material was injected into the clamp to reduce leakage. The inspector verified that the refrigerator was in place and that the Fermanite 2X material to be used was receipt inspected.
A Material Non-conformance Report (MNCR) was written on the missed receipt inspection of the clamp. The MNCR was closed based on the vendor's certi-ficate of compliance, an engineering evaluation of the clamp design cal-culation, and a QC inspection of the clamp during Fermanite inje:; tion.
The Maintenance Department also initiated a Quality Deficiency Report (QDR) against themselves for not identifying the missed receipt inspection and for not properly documenting it. Proposed corrective action included adding a sign-off, at a specific step in the job orders controlling con-tractor personnel, to verify receipt inspection.
The lic & a e verified that since the beginning of 1989 out of seven jobs invoiv'
r.ak Repair, Inc., only one, during February 1990, had missed the ren % inspection.
Based on this, the inspector concluded that this event was of isolated nature, and did not reflect a breakdown in the licensee's QA program with' Leak Repair, Inc. The licensee's corrective action was appropriate and acceptable.
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7.2 Safety Evaluation for Two Control Rods Moving at One Time Inspection 50-219/89-29 reviewed an event where two control rods were inadvertently selected and moved at the same time.
The design of the reactor manual control system enabled, through multiple manipulations of the rod select switches, the operator to select more than one control rod.
To address this concern the licensee implemented compensatory measures to require a second licensed operator be stationed to verify that only one control rod was selected and moved at a time. The licensee also identified that a more detailed review of the design and operation of the reactor manual control system rod select switches and interrupt circuitry was required.
Subsequent licensee review of the vulnerabilities of the reactor manual control system was documented in safety evaluation #644-001.
This evaluation reviewed the interrelationships and effects on safety of the reactor manual control system, rod worth minimizer, and the reactor protection system. The evaluation considered the effects from single malfunction or operator error of the reactor manual control system.
The safety evaluation identified the need for additional compensatory measures to protect against two control rods being moved at one time, in addition to stationing a second licensed operator to verify that only one rod is selected and moving.
For the first rod to be withdrawn, operators were to move the rod one notch and verify no other control rods moved.
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This verification establishes that no malfunction exists or goes undetected at the beginning of the startup.
The inspector concluded that ' licensee conipensatory measures provided additional assurance that only one control rod will be selected and moved at a time.
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The safety evaluation considered uncontrolled rod withdrawal from a suberitical or low power condition, uncontrolled rod withdrawal at power, and a control rod drop analysis. The evaluation concluded that the continued operation of Oyster Creek with the identification of a previously unanalyzed event for the initiation of a reactivity and power distribution anomaly does not constitute an unreviewed safety question.
This evaluation is limited to cycle 12 operation.
Further evaluation is required prior to moving fuel or control rods with the reactor head removed from the vessel.
Licensee completion of this evaluation and NRC review of the evaluation will be carried as an unresolved item.
(50-219/90-07-02)
7.3 Temporary Instruction 2500/27 NRC Bulletin 87-02 requested the licensee to test safety related and non-safety related fasteners to verify the fastenert met the requisite specifications.
This temporary instruction addressed the adequacy of the licensee's root cause analysis and corrective action to fasteners which were tested in response to Bulletin 87-02 and did not meet requisite specifications.
The licensee tested 20 safety related fasteners and 21 non-safety related fasteners in response to NRC Bulletin 87-02.
The results of these tests identified one safety related and four non-safety related fasteners that were out of specification.
The one safety related fastener (0C-002) had exceeded the specification for hardness.
The specification was RB 100. The tested hardness was RB 106. All other parameters tested were within acceptable limits, including the tensile strength and yield strength.
The licensee evaluated the use of the fasteners from this lot and determined that some had been used in non-safety _related applications.
The balance of the batch was not used and subsequently discarded.
No fasteners were used in safety related applications or in an ASME Code,Section III application.
Inspection Report 50-219/89-10 documented NRC review of the licensee's evaluation and disposition of the out of specification fastener.
The safety related fastener which failed the hardness test was procured in 1985. At that time, implementation of a more extensive inspection program was in progress.
This program was not fully in place until 1986.
The I
l program requires testing of material for hardness, a receipt inspection
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and visual inspection for all batches. Additionally, the licensee has in place a QA' vendor inspection plan where vendors are audited periodically, l
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14 A quality verification sample check, consisting of more extensive testing of material, is conducted on one vendor per year, The licensee reviewed the four out-of-specification, non-safety related fasteners. The licensee identified the locations where the fasteners were used and determined them to be acceptable. The unused fasteners were discarded.
Inspection Report 50-219/89-10 documented NRC review of the licensee's evaluation and disposition of the out-of-specification fasteners.
The non-safety related fasteners tested were classified as non-QA.
For installation of any of these fasteners into an application within the scope of the Ouality Assurance Program, the fasteners would require eval-untion and documentation to upgrade the fasteners. The licensee addi-tionally performs sampling of batches and documents deficiencies under Material Nonconformance Reports (MNCR).
In addition to corrective actions and evaluation resulting from MNCRs, the MNCRs are trended to assist in evaluation of the failures.
The licensee's corrective actions for the out-of specification fasteners were adequate.
Programs in place since 1986 provide greater assurance that out-of-specification safety-related fasteners are identified prior to use.
The out-of-specification safety-related fastener was procured in 1985 prior t'o full implementation of this program. Non-QA fasteners are tested and upgraded prior to use in a QA application.
The evaluation and resolution of the out-of-specification fasteners were appropriate.
None of these fasteners were used in either a safety related or ASME Code,Section III application.
7.4 Management Meeting A management meeting was conducted on April 6, 1990, in the Region I office.
The purpose of the meeting was to discuss recent performance in the areas of plant operations and radiological controls and to oiscuss the progress of site performance improvement initiatives.
The licensee presented their efforts in:
professionalism training;
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expectations and standards; and,
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team building.
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Two bargaining unit arbitrations were settled at the site. Also, an agreement was signed by site management with the bargaining unit establishing an equipment operator progression path to licensed control room operato IL o -
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operators via the Operator Concern Program and improved communications overall.
Radiological Controls management presented:
a summary of improvement plans;
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a summary of dose reduction efforts;
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several radiological areas showing improved short term trends; and,
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radiation protection staffing improvements.
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A copy of the licensee's presentation materials is enclosed as Attachment 11 to this report.
8.0 Inspection Hours Summary Inspection consisted of 203 direct inspection hours; 42 of these direct inspection hours were performed during backshift periods, and 21 of these hours were deep backshift inspection.
9.0 Exit Meetings and Unresolv'ed Items 9.1 Preliminary Inspection Findings
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A verbal summary of preliminary findings was provided to the senior licensee management at the conclusion of this inspection.
During the inspection, licensee management was periodically notified verbally of the preliminary findings by the resident inspectors. No written inspection material was provided to the. licensee during the inspection.
No proprietary information is included in this report.
9.2 Unresolved Items
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Unresolved items are matters for which more information is required in order to ascertain whether they are acceptable, violations or deviations.
Unresolved items are discussed in paragraphs 4.2 and 7.2 of this report.
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ATTACHMENT I Personnel Contacted Licensee Personnel F. Barbieri, Mechanical Systems Engineer
- J. Barton, Dyster Creek Deputy Director R. Barrett, Plant Operations Director J. Boyle, Plant Operations, GOS R. Brown, Radwaste Operations Manager
- G. Busch, Licensing Manager P. Cervanka, Plant Operations Engineering G. Cropper, Operatory Training Manager P. Crosby, Plant Engineering
- B. De Merchant, Licensing Engineer
"R.
Fenti, QA Mod /0PS Mgr.
E. Fitzpatrick, Vice President & Director V. Foglia, Technical Functions Manager D. Jerko, Licensing & Reg. Affairs D. Jones, Plant Engineering J. Knubel, Nuclear Security J. Kowalski, Manager Training
"L. Lammers, Plant Maintenance Director
- D. Leroy, Electrical Maintenance Supervisor K. Mulligan, Plant Operations R. Perry, Radiological Engineer D. Ranft, Engineering Manager J. Rogers, Licensing Engineer
- A. Rone, Plant Engineering Director P. Scallon, Plant Operations Manager
P. Tanburro, Mechanical Systems Engineer D. Tuttle, Radiological Controls Deputy Director J. Ventosa, Plant Engineering NRC Personnel
- M. Banerjee
- E. Collins
Denotes attendance at exit meetin _
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ATTACHMENT II 4/6/90 Management Meeting with GPU Nuclear Attendees GPUN M. J. Slobodien, Radiological Controls Director J. E. Hildebrand, Radiological Controls Director R. J. Barrett, Plant Operations Director D. E. Tuttle, Sr., Radiological Controls Deputy Director E. E. Fitzpatrick, Vice President / Director J. L. Sullivan, Jr., Licensing / Regulatory Affairs Director J. J. Rogers, Licensing Manager (Acting)
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M. Banerjee, Resident Inspector, OC D. Beaulieu, Reactor Engineer, DRP D. Tondi, Acting Section Chief, DRP E. Collins, Senior Resident Inspector, OC T. Dragoun, Acting Section Chief, FRPS S. Sherbini, Radiation Specialist, DRSS D. Lew, Resident Inspector, OC W. Kane, Director, DRP R. Bellamy, Branch Chief, DRSS E. Wenzinger, Branch Chief, DRP-i
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I MANAGEMENT MEETING L
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d Oyster Creek Nuclear Generating Station April 6, 1990 I
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AGENDA INTRODUCTION E. E. FITZPATRICK OPERATIONS R. J. BARRETF RADIOLOGICAL PERFORMANCE l
IMPROVEMENTSD D.E. TUTTLE
!I RADIOLOGICAL CONTROLS DEPARTMENT M. J. SLOBODIEN
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ORGANIZATION CHANGES I
TEAMBUILDING EFFORTS
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RESOLUTION OF LONG STANDING BARGAINING UNIT ITEMS
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I ORGANIZATION CHANGES j
I REASSIGNMENT OF MCF DIVISION FUNCTIONS
PLANT DIVISION CONTROLS MAINTENANCE
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ORGANIZATION CHANGES (CONTD)
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REASSIGNMENT OF MCF DIVISION FUNCTIONS
I PLANT DIVISION
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NEW SITE SERVICES DIVISION
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NEW SITE SERVICES DIVISION CONSTRUCTION
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FACILITIES OUTSIDE PROTECTED AREA
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ADMINISTRATION OF WORK MANAGEMENT
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SYSTEM AND INTEGRATED SCHEDULING PROCESS
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PLUS OTHER
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PLANT DIVISION
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MAINTENANCE
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OUTAGE MANAGEMENT INCLUDING INTEGRATED
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SCHEDULE CONTROL
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ORGANIZATION CHANGES (COhTD)
I INTERNAL RADIOLOGICAL COS"IROLS DEPARTMEh"I'
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MAN ALARA GROUP FROM RAD ENGINTERING
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GROUP REPLACE WITH SEVERAL HP PROFESSIONALS
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TEAMBUILDING EFFORTS I
EXPECTATIONS AND STANDARDS
K KDfOLOGICAL PERydRMANCE COMMITTEE
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MODEL STANDARD
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HEAT STRESS, FALL, LIFTING
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PROCEDURES: WRITERS' STANDARD,.
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PROCEDURE COMPLIANCE, SELF CHECKING
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HOUSEKEEPING: MODEL AREAS, INSPECTION
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GUIDELINES INCLUDING MATERIEL CONDITION, J
INDUSTRIAL SAFETY, RAD PROTECTION,
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CLEANLINESS STANDARDS, PAINTING i
STANDARD, LABELING STANDARD
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EXCELLENT PARTICIPATION / ACCEPTANCE
' LOCAL COMMUNITY BORROWED AND IMPLEMENTED
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CONCERNS-l
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EFFORT: TASK TEAM OF 5 B/U AND 4 MANAGEMENT PERSONNEL
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TEAMBUILDING TRAINING INITIATIVES COMPLETED I
ALMOST ALL MANAGERS /
3-4 DAYS 1987 88 SUPERVISORS
ERROR FREE REFUEL-12 HOURS 12/88 OPERATORS, OPERATIONS l
MANAGEMENT, CORE ENGINTERING I
COMMUNICATIONS DEPARTMENT 3 DAYS 4/89,7/89
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NEW MANAGERS / SUPERVISORS 3 DAYS 6/89 (2)
8/89 I
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MANAGEMENT TEAM 1 DAY 9/89 I
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RADCON DEPARTMENT 1 DAY 11/89 l
PLANT ENGINEERING / PLANT 1 DAY 1/90
MATERIEL / TECH FUNCTIONS
QA DEPARTMENT 3 DAYS 1/90 (2)
CHEMISTRY SUPPORT 3 DAYS 1/90,3/90
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DEMING SEMINARS 2 DAYS 7/89,11/89 (MANAGERS AND 1/90,3/90
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SCHEDULED I
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SAFETY COMMITTEE 3 DAYS 5/90 I
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(LICENSING, HUMAN RESOURCES)
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TRAINING DEPARTMENT 3 DAYS 8/90
OUTAGE WORKERS /
3 DAYS 9/90 SUPERVISORS
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OUTAGE WORKERS /
3 DAYS 10/90 SUPERVISORS
PLANT MAINTENANCE 3 DAYS 10/90 I
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SCHEDULED
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DEPARTMENT PLANT OPERATIONS / PLANT 3 DAYS 11/90
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DEMING SEMINARS 2 DAYS LATE 1990
(MANAGERS / SUPERVISORS)
' OBSERVATION TECHNIQUES 1 DAY 1/90 (2)
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5/90 (2)
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ADDITIONAL TEAMBUILDING EFFORTS ESTABLISHED NEW RADIATION PERFORMANCE COMMI1 TEE
COLLABORATIVE B/U AND MANAGEMENT
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AND OVER 200 DAYS)
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AWAITING FINAL SIGNATURE
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ESTABLISHES ACADEMIC REVIEW BOARD
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ADVANCEMENT ISSUES I~:
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Lg; PERSONNEL DEVELOPMENT INITIATIVES INPO EVALUATION OBSERVERS / PEERS (1989)
5 SENIOR MANAGERS I
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SCHEDULING ADDITIONAL MANAGERS IN 1990
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DEPUTY DIRECTOR 1989
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PLANT ENGINEERING DIRECTOR - 1990 -
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ROTATED OPERATIONS MANAGER AND RADWASTE
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MANAGER FROM TECHNICAL FUNCTIONS PLANT ENGINEERING SENIOR MANAGER TO
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LICENSE TRAINING
46 IN UNIVERSITY OF MARYLAND DEGREE PROGRAM
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2 SRO'S IN FULL TIME DEGREE PROGRAM
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PROCEDURES ARE TECHNICALLY SOUND
PROCEDURES ARE HUMAN FACTORED
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ACTIONS IN PROGRESS
PROCEDURE WRITERS' STANDARD UNDER DEVELOPMENT (SURVEILLANCE, OPERATING, ADMINISTRATIVE PROCEDURES)
MODEL PROCEDURES 11/07/89
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COMPLETED
FIRST DRAFT STANDARD 12/07/89 I
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SECOND DRAFT STANDARD 02/01/90
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FINAL STANDARD 05/90
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ERDCEDURE COMPLIANCE (CONT'D)
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PROCEDURE UPGRADE PROGRAM CORE SPRAY SYSTEM (PILOT)
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OPERATING AND SURVEILLANCE PROCEDURES WILL
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BY THE FOURTH QUARTER 1990.
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ALL NEW PROCEDURES AND THOSE WITH CHANGES
-
ENCOMPASSING 50% OF THE PROCEDURE WILL BE UPGRADED TO THE WRITERS' STANDARD
!l HUMAN PERFORMANCE UPGRADE PROGRAM
i J
i INCLUDES SELF CHECKING TRAINING
-
(I&C, MECH, ELEC, CHEM, LICENSED /
i
,
NON LICENSED OPERATORS)
I
,
.
INCLUDES PERIODIC REVIEW OF COMPANY /
-
SITE POLICY i
,
,:
FUTURE ACTIONS
INTEGRATE MAINTENANCE PROCEDURES WRITERS' GUIDE l
INTO OYSTER CREEK WRITERS' STANDARD
,
.
REVISE EXISTING PROCEDURES TO BE CONSISTENT
L WITH WRITERS' STANDARD (SCHEDULED PENDING 4-COMPLETION OF PILOT)
Lg:
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I PROCEDURE USE AND ADHERENCE I
COMPARED CURRENT OYSTER CREEK REQUIREMENTS /
GUIDANCE TO DRAFT INPO GOOD PRACI' ICE CURRENT OYSTER CREEK REQUIREMENTS /
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GUIDANCE ARE CONSISTENT / EXCEED INPO
,
GOOD PRACTICE GUIDANCE g
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PL ANT OPER ATIONS
- STAFFING
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- COMMUNICATIONS
.
- PROCEDURES (
- LABELING
' TRAINING / QUALIFICATIONS
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SIX SHIR ROTATION ACHIEVED ON FEBRUARY 5,1990 I
'IWO (2) NEW GROUP SHIR SUPERVISOP.S j
-
'IWO (2) SROs TRANSFERRED 'IO OPERATIONS AS GOSs
-
.
-
l ELEVEN (11) CROs LICENSED AND QUALIFIED
-
NINE (9) NEW EQUIPMENT OPERATORS
-
,
ONE (1) NEW RADWASTE SHIR SUPERVISOR
-
NINE (9) NEW RADWASTE OPERATORS
-
I
,
,
SRO LICENSED TRAINING COORDINATOR ESTABLISHED MID 1989 l
i
-
ESTABLISH DAY SHIR GROUP SHIR SUPERVISOR POSITION
i (MID 1990)
I
ESTABLISH ROTATIONAL PROGRAM FOR SELECTED SROs INTO KEY POSITIONS OUTSIDE OPERATIONS (MID 1990)
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e S T A F F I N G (CONTINUED)
PIPELINE CONTINUATION L
NEW HOT LICENSE CLASS STARTED 3/5/90
-
SIX (6) CONTROL ROOM OPERATORS (ROs)
DVO (2) OPERATIONS STAFF ENGINEERS (SROs)
FOUR (4) OTHER DEPARTMENT (SROs)
l TWO (2) CROs - TO BE PROMOTED TO GOS
-
NEW NLO CLASS TO STARTIN MAY 1990
-
-
OPERATOR RESIDENT STUDENT PROGRAM EMPLOYEE ATTENDS COLLEGE FULL TIME
-
EMPLOYEE EARNS BACHELOR'S DEGREE
-
DEGREE MUST BE TECHNICAL (ENGINEERING, ETC.)
-
TWO (2) SRO LICENSED GOSs CURRENTLY ENROLLED
-
L EXPECT TO ENROLL BVO (2) MORE EMPLOYEES IN FALL '90
-
l FIRST GRADUATE MAY '90
-
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CO M MUNIC ATIONS
.
OPERATOR CONCERN PROGRAM
ESTABLISHED 5/14/89
=
-
274 CONCERNS SUBMITTED
-
198 CONCERNS CLOSED
-
WIDESPREAD ACCEPTANCE
!
-
USED BY SROs, CROs, NLos, CHEMISTRY & OTHERS
-
,
WIDE CATEGORY DISTRIBUTION *
-
DRAWINGS / PROCEDURES 2 h'
EQUIPMENT / HARDWARE
- HUMAN / PEOPLE ITEMS
OTHER
- SOME ITEMS FALL INTO 1WQ OR MORE CATEGORIES.
,
l
DEVIATION REPORT SYSTEM SYSTEM IMPROVED
-
EMPLOYEE AWARENESS
-
THRESHOLD FOR USE LOWERED l
-
ROOT CAUSE DETERMINATION L
-
'
'
VERBAL COMMUNICATIONS l
PROCEDURE REVISED
-
EMPHASIZED IN SIMULATOR
-
REPEAT BACK OF ORDERS
-
ERROR FREE PHILOSOPHY
-
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L C O M M U N I C A T I O N S (CONTINUED)
CERTIFICATIONS i
.
FOR ALL MILESTONE EVOLUTIONS j
.
-
RESTART
REFUEL
ETC.
MANAGEMENT REVIEW AND APPROVAL
-
SUPPORTS ERROR FREE PHILOSOPHY
-
-
ACTION PLANS FOR ALL MAJOR EVOLUTIONS
-
POWER ASCENSION
POWER DESCENSION
DRYWELL INSPECTIONS
.
SIGNIFICANT SYSTEM OPERATIONS
,
COORDINATION OF DEPARTMENTS
-
MANAGEMENT COVERAGE DURING STARTUPS
-
'
ERROR FREE PHILOSOPHY
-
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PROCEDURES
.
l.
SYSTEM AND OPERATIONS SURVEILLANCE PROCEDURES
REASSIGNED TO OPERATIONS
'
MOST NOW REVIEWED BY LICENSED PERSONNEL
-
'
STAFF ENGINEER TRACKS REVIEW / REVISION PROCESS
OVERDUE REVIEWS VIRTUALLY ELIMINATED (1 OCCURRENCE)
-
MULTIPLE CHANGES COMBINED INTO ONE
-
f
QUALITY OF REVIEWS IMPROVED
STANDARD REVIEW CHECKLISTS DEVELOPED l
-
i REVIEWER COMPLETES CHECKLIST FOR EACH REVIEW
-
DIFFERENT REVIEW REQUIREMENTS FOR DIFFERENT
-
TYPES OF PROCEDURES.
LESSONS LEARNED INCORPORATED INTO CHECKLISTS
-
PROCEDURE IN HAND REQUIREMENTS
--
PROCEDURALLY DEFINED
~
PROCEDURES STAGED THROUGHOUT STATION
-
l
'
'
WALK THROUGH DETERMINATION i
MAJOR REVISIONS
-
NEW PROCEDURES
-
DO IT RIGHT THE FIRST TIME (ERROR FREE)
-
.
.
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'-
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PL ANT L ABELING I.
OVERVIEW
SYSTEMS TO BE LABELED 112
-
TOTAL NUMBER OF LABELS 60,000 l
-
TO COMPLETE 75% BY 5/1/91
-
,-
j CURRENT STATUS
SYSTEMS NO. LABELS COMPLETED 12 (11 %)
5,480 ( 9 %)
q
..
.
IN PROCESS 9 ( 8%)
5,950 (10 %)
!
l
-
OPERATOR OWNERSHIP
!
!
ONE (1) CRO FULL TIME
-
ONE (1) CHEMISTRY TECHNICIAN FULL TIME q
-
DVO (2) RADWASTE OPERATORS FULL TIME
-
EXTRA DAY SHIFT SUPPORT AS NEEDED
-
l PLAN TO ASSIGN ONE (1) EO FULL TIME BY 5/1/90
--
BENEFITS
'
MINIMIZE MISTAKES (OPERATIONS & MAINTENANCE)
-
REDUCED RADIATION EXPOSURE (ALL EMPLOYEES)
-
~
PROCEDURE IMPROVEMENTS
-
DRAWING IMPROVEMENTS
-
COMPUTERIZED ROUNDS (FUTURE)
-
COMPUTERIZED LINEUPS (UNDER DEVELOPMENT)
-
,
I
-
- -
-
.
.
,
-
TRAINING /OUAf iFICATIONS -
GROUP SHIFT SUPERVISOR QUALIFICATION PROGRAM (NEW)
p
'I 3 - 6 MONTHS DURATION -
l
-
{
'
SUPERVISORY / MANAGEMENT SKILLS
{
-
STAND WATCHES IN TRAINING
-
INCLUDES SUCCESSFUL SHIFT MANAGEMENT TRAINING
-
,
GROUP OPERATING SUPERVISOR QUALIFICATION PROGRAM (NEW)
6-9 MONTHS DURATION i
-
OBTAIN SRO LICENSE
-
-
SUPERVISORY / MANAGEMENT SKILLE
-
STAND WATCHES IN TRAINING i
-
INCLUDES SUCCESSFUL SHIFT MANAGEMENT TRAINING
-
,
'MAC CORPORATION SUCCESSFUL SHIFT MANAGEMENT TRAINING
FIVE (5) DAY PROGRAM
-
FOUR (4) SUPERVISORY PERSONNEL TRAINED PRE 1989
-
TEN (10) SUPERVISORY PERSONNEL TRAINED IN 1989
-
._
TEN (10) SUPERVISORY PERSONNEL SCHEDULED FOR JUNE '90
-
I
INPO HPES EVALUATOR TRAINING
.
PRESENTED BY INPO (3 DAY PROGRAM)
-
--
-
INTEGRATED WITH IMPROVED CRITIQUE PROCESS I
-
NINE (9) OPERATIONS DEPARTMENT PERSONNEL TRAINED
-
NINE (9) ADDITIONALTO A'ITEND IN 1990
-
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TRAINING /OUALIFICATIONS (CONTIN 0EDJ
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REFUELTEAM TRAINING
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FIVE (5) DAY PROGRAM
.
-
AT GE's SAN JOSE FACILITY
-
.
THREE (3) TEAMS OF SEVEN (7) (21 OPERATORS TOTAL)
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OPERATOR PROGRESSION L
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CROs MUST BE PRIOR NLOs r
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UNION AGREEMENT PENDING
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QUALITY IMPROVED
-
WORKER PARTICIPATION
-
TIMELINESS IMPROVED
-
ACTIONS 1 RACKED TO COMPLETION
-
UTILIZES INPO'S HPES APPROACH
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EQUIPMENT MAINTENANCE PRIORITIES
OPERATIONS CHAIRS P.O.D.. SETS DAILY PRIORITIES
-
l OPERATIONS PRIORITIZES ALL MAINTENANCE REQUESTS
.
FULL DECK EMPHASIS
,
-
.
.
- I
MODIFICATIONS TO PLANT OPERATIONS NOW OWNS TURNOVER PROCESS
'
.
I TIE IN APPROACH EMPHASIZED
-
I EQUIPMENT CONTROL l
'
'
COMPUTERIZED TAGGING IMPLEMENTED
-
STANDARDIZED TAGOUTS IN COMPUTER
-
'
SUPPORTS ERROR FREE CONCEPT
-
I
.
OPERATIONS DEPARTMENT PROJECIS
'
FUEL POOL CLEANUP (COMPLETED ON SCHEDULE)
.
,
240,000 CURIES - INCLUDED 76 CONTROL RODS,
'
5 FUEL CHANNELS,135 POISON CURTAINS,25 LPRMS, ETC.
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CRD DISPOSAL PROJECT (COMPLETED ON SCHEDULE)
-
l 2 CURIES. INCLUDED 58 CRDs
REFUEL BRIDGE REPLACEMENT (ERROR FREE CONCEPT)
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HOT SPOT / LOCKED DOOR REDUCTION - AIARA
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E O U I P M E N T (CONT.:NUED)
- NOTEWORTHY RESULTS AOG OPERABILITY
-
99% THIS CYCLE VS 90% PRIOR CYCLE REACTOR WATER QUALITY
-
0.082 us/cm THIS CYCLE VS 0.110 us/cm PRIOR CYCLE
-
AEOG NOBLE GAS RELEASE RATE
-
11,500 ucl/sec THIS CYCLE VS 40,000 ucl/sec PRIOR CYCLE STACK GAS RATE S MDA WITH AOG IN SERVICE
,
OVERBOARD RELEASES
.
5.5% IN 1989 (8,978,588 GALS. PROCESSED)
0% IN 1990 (2,645,672 GALS. PROCESSED)
DEMINERALIZER PROCESSING SYSTEM BEING INSTALLED
"
LCO RATE THIS CYCLE
.
05/16/89 07/11/89 4.36 DAYS / WEEK 07/12/89 09/10/89 3.11 DAYSAVEEK 09/11/89 11/10/89 4.86 DAYS / WEEK 11/11/89 01/11/90 2.28 DAYSAVEEK
01/12/90 03/08/90 1.78 DAYSAVEEK 03/09/90 04/02/90 0.58 DAYSAVEEK
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6'-
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RADIOLOGICAL PERFORMANCE TASK FORCE
'
.
ALL RECOMMENDATIONS PIACED ON OC PLAN FOR EXCELLENCE
.I ALL RECOMMENDATIONS INCLUDED IN RADIOLOGICAL
-
IMPROVEMENT PIAN
RADIOLOGICAL IMPROVEMENT PLAN COMMITTEE ESTABLISHED OCTOBER 1989
.
MEMBERS: VARIOUS PLANT AND SUPPORT DIVISION
.
,
REPRESENTATIVES
!
FUNCTIONS: CONSOLIDATE AND PRIORITIZE ALL l
'
.
ACTION ITEMS, INITIATIVES, RECOMMENDATIONS
'
AND COMMITMENTS INTO A CONSOLIDATED PIAN PLACED ITEMS IN OC PIAN FOR EXCELLENCE
.
STATUS: 95 ITEMS PLACED IN PLAN
-
I 47 ITEMS COMPLETED 48 ITEMS OPEN
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,g RADIOLOGICAL IMPROVEMENT PLAN UPDATE I
.
RADIOLOGICAL PERFORMANCE TASK FORCE i
RADIOLOGICAL IMPROVEMENT PLAN
COLLECTIVE DOSE REDUCTION
IMPROVED RADIOLOGICAL PERFORMANCE INDICATORS
'
ACTIONS UNDER DEVELOPMENT TO FURTHER IMPROVE
RADIOLOGICAL INDICATORS
TREND ANALYSIS I
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i PRINCIPLE SOURCES OF ACTION ITEMS I
TDR 941, REY. O. EXPOSURE REDUCrlON PLAN, OCNGS (AUGUST 1988),
I l
.
INPO ASSIST VISIT (NOVEMBER 1988),
TDR 941, REY.1. EXPOSURE REDUCTION PLAN, OCNGS (APRIL 1989),
INPO OUTAGE hiANAGEMENT ASSIST VISIT _(51 ARCH 1989),
i
NRC SALP REPORT (htAY 1989)
I i
INPO EVALUATION AND ASSISTANCE VISIT (JUNE 1989),
OC PLAN FOR EXCELLENCE IN PERFORh1ANCE (JUNE 1989),
RADIOLOGICAL CONTROLS PERFORMANCE TASK FORCE REPORT (JULY 1989).
I
'
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CATEGORIES OF ITEhtS
'
j OPEN COh1PLETE
.
SOURCE TERht/ DOSE RA.'E REDUCTION
7
,
Ihf PROVED MGhfT,/ SUPERVISORY INVOLVEhfENT
15
I
Ih1 PROVED WORKER PERFORhiANCE
8
,
.
Ih1 PROVED WORK P!ANNING
4 ll
'
Ih1 PROVED WORK ENVIRONh1ENT
3
'
REDUCI' ION IN WORK
7
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Ih1 PROVED FACILITIES / TOOLS
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TOTALS
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EXPEL 7ATIONS iI SHORT TERM SHORT TERM EXPELTATIONS ARE DIRECTED TOWARD -
I IMPROVED IMPLEMENTATION OF THE EXISTING RADIOLOGICAL CONTROLS PROGRAM BY THE SITE WORK FORCE.
I I
LONG TERM LONG TERM EXPECTATIONS ARE DIRECTED TOWARD WORK FORCE PERFORMANCE WITH HEAVY EMPHASIS I
TOWARD PLANT SOURCE TERM REDUCTION TO SUPPORT OUTAGE WORK.
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COI I FCTIVE DOSE REDUCTION ITEMS
'
THREE STEP CHEMICAL DECON OF RECIRCULATION AND i
REACTOR WATER CLEAN UP SYSTEM l
COBALT ELIMINATION / REDUCTION
'
!
25 CRD BLADES 13R
INCREASED RWCU FLOW AFTER 13R L.P. TURBINE BLADE REPLACEMENT i
UTILIZATION OF VIDEO EQUIPMENT
. VIDEO DISC SYSTEM VIDEO CAMERA IN CONDENSER / HEATER BAY
- l- - I
. VIDEO CAMERAS IN DRYWELL 13R l
. VIDEO CAMERAS IN Hl/LO CONDUCTIVITY STEAM JET AIR EJECTOR ROOM
RAD WASTE PROCESSING SYSTEM ALTERNATIVES NEW FUEL HANDLING BRIDGE
HOT SPOT ELIMINATION
i HYDROIAZE R.B, FLOOR & HUB DRAINS Ig PREPARE FOR HYDROLAZE RBEDT g
. CRD REBUILD ROOM HOT SPOTS
. PLANT OPERATIONS IS LEAD ON PROGRAM PLANT DECONTAMINATION PROGRAM
'
. DEVELOPED AND MANAGED BY PLANT MATERIEL
PLANT MODIFICATIONS TO REDUCE DOSE
. RESIN ADDITION PIPE. COMPLETED
- E REDUCE DWEDT DOSE BY MINIMIZING SLUDGE B
INTRODUCTION FOLLOWING REFUELING OPERATIONS
. DRYWELL IMPROVEMENT PIAN
<
I
FUEL POOL EQUIPMENT CLEAN UP I
240,000 CURIES IN 10 MONTHS 16'B" TYPE SHIPMENTS NO PROBLEMS
HOUSEKEEPING IMPROVEMENTS
. CRD REBUILD ROOM CLEAN UP
,
i-NORTH EAST CORNER ROOM CLEAN UP l -
T.B. BASEMENT CLEAN UP BEHIND LHRA DOORS I
.
.,_-,_..._.---..--_.,,..,__...---,_._.-...-_.___-m
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DRnVELL EXPOSURE REDUCTION PROGRAM SCOPE i
PROGRAM CONTAINS FORTY (40) SEPARATE PROJECTS [fASKS PROJECTS WILL REDUCE EXPOSURE BY VARIOUS MEANS:
WORK ELIMINATION
IMPROVEMENT OF EXISTING SERVICES
'
,
'
SOURCE TERM REDUCTION
IMPROVING EQUIPMENT RELIABILITY
SHIELDING
'
INCREASING PRODUCTIVITY I
PROVIDING BETTER TOOLING
'
IMPROVING COMMUNICATIONS l
PERSON REM SAVINGS
$
TOTAL SAVINGS PER OUTAGE * AFTER ALL PROJECTS ARE IMPLEMENTED,
'
ARE ESTIMATED TO RANGE FROM 110185 PERSON. REM i
- DOES NOT INCLUDE PERSON. REM COSTS TO IMPLEMENT
'
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I l
'
'
NO LOCKED HIGH RADIATION AREA VIOLATIONS SINCE 1 AUGUST 1989
DEPARTMENT MANAGERS DEVELOPED RADIOLOGICAL TARGETS
'
OPERATORS USE OF VIDEO SYSTEMS
RECIRCULATION PUMP SEAL WORK IN FEBRUARY 1990 DECONTAMINATION OF SEAL PRIOR TO REBUILD
.
MOCK UP UTILIZATION FOR SEAL REMOVAL /
.
INSTALLATION
'I
'
RECIRCULATION PUMP SEAL WORK IN MARCH 1990 I
-
.
EXTENSIVE RADIOLOGICAL WORK PLANNING DECONTAMINATION OF SEAL PRIOR TO REBUILD
.
MOCK UP UTILIZATION
-
OVERALL RADIOLOGICAL PERFORMANCE IMPROVED
-
'
OVER PAST PERFORMANCE
SUGGESTION TO MINIMlZE DOSE DURING RESIN I.
ADDITION
'E WORKER TAKING PROPER ACTION WHEN DOSE RATES WERE
E HIGHER THAN EXPECTED IN FEBRUARY 1990
'
12UJ OUTAGE
'
BETTER WORK PLANNING
-
WORK MET SCHEDULE
-
-
TOTAL COLLECTIVE DOSE WAS FAR LESS TilAN
-
SIMILAR SCOPE OUTAGES I
I
.
.-
.
.
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-
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AcrIONS UNDERWAY TO FURTHER IMPROVE RADIOLOGICAL PERFORMANCE I
FORMA'110N OF RADIOLOGICAL PERFORMANCE COMMITTEE
.
ESTABLISHED ADVANCED RADIATION WORKER TRAINING
-
ONGOING REVIEW OF PROTECTIVE CLOTHING IN USE
.
,
-E, TREND ANALYSIS
{
.'
u CONTAMINATED AREA REDUCTION
-
LOCKED HIGH RAD AREA VIOLATION REDUCTION
-
,
REDUCED PERSON REM j.
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REDUCED SKIN & CLOTHING CONTAMINATION
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RADIOLOGICAL ENGINEERS ALARA SPECIALIST RADIOLOGICAL ElNINEER
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