IR 05000244/1989014
| ML17250B032 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 10/20/1989 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17250B031 | List: |
| References | |
| 50-244-89-14, NUDOCS 8911020227 | |
| Download: ML17250B032 (22) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-244/89-14 Licensee No.
DPR-18 Licensee:
Rochester Gas and Electric Corporation 49 East Avenue Rochester, New York Faci 1 ity:
Location:
R.
E. Ginna Nuclear Power Plant Ontario, New York Inspection Conducted:
September 11 through October 10, 1989 Inspectors:
Approved by:
N. Perry, Acting Senior Resident Inspector, Ginna E. Yachimiak, Acting Resident -Inspector, Ginna L.
E.
C. McCabe, Chief, Reactor Projects Section 3B
~I20(
Date
~Summa r Areas Ins ected:
Routine resident inspection of station activities including plant operations,,radiological controls, maintenance, surveillance, security, periodic and special reports, self-assessment capabilities, and fire brigade shift manning.
Inspection activities consisted of 124 hours0.00144 days <br />0.0344 hours <br />2.050265e-4 weeks <br />4.7182e-5 months <br /> of inspection, including 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of backshift inspection and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of deep backshift in-spection on September 17, 1989, and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of deep backshift inspection on September 24, 1989.
Results:
Overall, the plant operated safely during this inspection period.
Plant events were dealt with effectively.
Good ALARA planning of maintenance activities was noted (section 3.c)
as was good control of. activities during the first open house (section 3.e).
A discrepancy was identified in the licensee's fire brigade composition (section 3.h)
and was promptly corrected.
89fl020227 ee]020 PDR ADOCK 0,.0002M
TABLE OF CONTENTS PAGE 1;
Persons Contacted.
2.
Summary of Plant Operations......
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3.
Functional or Program Areas Inspected.
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a
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b.
C.
d.
e.f.
g.
h.
Plant Operations (71707)..
Radiological Controls (71707).....,
Maintenance (62703)
Surveillance (61726).
Securi ty (71707).
Periodic and Special Reports (90713)
Evaluation of Licensee Self-Assessment Fire Brigade Shift Manning (71707).....
Capabilities (40500)..
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4.
Exit Interview (30703)...
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DETAILS Persons Contacted During this. inspection period, inspectors held discussions with and inter-viewed operators, technicians, engineers and supervisory level personnel.
The following people were among those contacted:
"S. Adams, Technical Manager
- T. Alexander, gC Engineer M. Cavanaugh, Fire Protection Engineer
- D. Fi lkins, Manager of HP & Chemistry
- N..Goodenough, Maintenance Technical Analyst G.. Hermes, Senior Licensing Engineer A. Jones, Corrective Action Coordinator
- M. Lilley, Nuclear Assurance Manager
- R. Marchionda, Director of Outage Planning T. Harlow, Superintendent, Support Services R. Mecredy, General Manager, Nuclear Production A. Morris, Maintenance Manager
- J. St. Hartin, Corrective Action Coordinator
- T. Schuler, Operations Manager L. Smith, Operations Supervisor
- S. Spector, Superintendent, Ginna Station
- J. Widay, Superintendent, Ginna Production
- R. Wood, Supervisor, Nuclear Security
- Present at exit meeting on October 10, 1989.
Summar of Plant 0 erations At both the beginning and close of the inspection period the plant was at approximately full power.
On September 20, 1989, the plant experienced a containment ventilation isolation; the isolation signal was generated by containment radiation monitor circuitry.
The isolation was verified by the operators, who noted that no plant computer system alarm was generated and that main control board radiation monitors indicated normal.
Subsequent troubleshooting by maintenance personnel found no failures or abnormalities.
A four-hour notification was made to the NRC.
The system was reset and returned to service; no further problems were experienced.
A turbine runback occurred on October 7, 1989, resulting in a
20 percent decrease in turbine power.
Corrective actions were taken and the plant was brought back to full power operatio '
The turbine runback occurred 'after control room operators had completed the gain adjustment for power range (PR) nuclear channel N-44 and returned its Dropped Rod Mode Selector Switch from the
"BYPASS" to the
"NORMAL" position.
The plant responded
- as designed and the operators performed the actions required by Abnormal Procedure (AP)-TURB.2, "Automatic Turbine Runback."
The suspect power range channel (N-44) was removed from service in accordance with Equipment Restor'ation procedure (ER)-NIS.3,
"PR Mal-function."
Troubleshooti'ng activities performed on the N-44 PR nuclear instrument channel did not identify any problems associated with the in-strument's performance.
However, it appeared that, since the power range dropped/rod stop 5 percent per 5 seconds Main Control Board (MCB) annun-ciator was lit, the rate of change setpoint for this turbine runback func-tion was exceeded.
The actuation of the power range dropped rod/rod stop bistable for this channel caused the turbine to runback to 80 percent power.
The root cause of this runback was not found:
All other required follow-up actions were taken and the appropriate notifications were made.
3.
Functional or Pro ram Areas Ins ected Plant 0 erations (71707)
The inspectors verified that the R.
E. Ginna Nuclear Power Plant operated safely and in conformance with license and regulatory re-quirements.
Portions of Rochester Gas and Electric Corporation's management control system were evaluated to ensure effective dis-charging of its responsibilities for safe operation.
Control room staffing was adequately maintained and access to the control room was properly controlled.
Operators adhered to approved procedures and understood the reasons for lighted annunciators.
Control room log books were reviewed to obtain information concerning trends and ac-tivities, and recorder traces were observed for abnormalities by the inspectors.
Shift turnovers were adequate to ensure necessary in-formation was addressed.
The inspectors verified that selected Engi-neered Safety Feature valves and breakers were correctly aligned, and portions of the containment isolation lineup were correct.
All ac-cessible areas of the plant were toured and plant conditions and ac-tivities in progress were observed by the inspectors with no inade-quacies identified.
The inspectors verified that plant technical specifications were complied with and selected safety-related tagouts were proper.
Diesel Generator 0 erations On September 20, 1989, the "A" Diesel Generator (D/G) monthly sur-veillance was performed.
The. test was progressing satisfactorily until the 0/G was stopped by the control room operator.
At this point, the control room operator turned the
"D/G A CONTROL" switch to the
"STOP" position, pushed the "D/G A VOLTAGE SHUTDOWN" button, and then pushed the "0/G A RESET" and
"D/G A FIELD RESET" buttons.
When the "0/G A RESET" button was pushed the D/G restarted and
trippe'd on overspeed.
The licensee attributes the cause of this re-start and overspeed to the operat'or not waiting for a full minute to elapse after turning the "D/G A CONTROL" switch.
By pushing the "0/G A RESET" switch while the D/G was still coasting down, the fuel racks realigned to their full open position and caused the 0/G to restart and trip on overspeed.
Subsequent to this series of events, the Control Room Foreman (CRF)
directed the qualified surveillance test operator reset the overspeed trip at the D/G locally.
This action enabled the control operator to push the "D/G A RESET" button again, and thus restore the D/G to standby status for automatic starting.
Realizing that the D/G may not have been properly reset after the termination of this surveil-lance test, the Shift Supervisor (SS) directed the Results
& Test (R&T) supervisor to modify the just completed surveillance procedure.
The resulting temporary surveillance procedure was reviewed by the SS and implemented by the R&T supervisor.
Instead of running the D/G for a full two hours, as previously done, the D/G was merely started and verified to have reached full speed.
After a few minutes the 0/G was stopped.
This time, the 0/G was verified as having come to a
full stop before the control room operator pushed the
"D/G A RESET" switch.
The D/G did not attempt to restart.
The Hechanic Foreman witnessed this surveillance and found its performance to be satis-factory.
This retest was deemed by the licensee as an acceptable means of assuring operability of the D/G.
The licensee concluded that personnel error was the probable cause of the 0/G restart and overspeed trip ~
However, the exact time be-tween stopping the 0/G and pushing the "0/G A RESET" button could not be verified.
The licensee initiated a procedural revision to require the 0/G to be locally verified as having come to a complete stop be-fore the control room operator pushes the "0/G A RESET" button.
The inspector concluded that these actions were appropriate.
Containment Entries On September 24, 1989, control room personnel noted an increase in the frequency of containment recirculation fan condensate draining operations
.and a decrease in time between cycling operations of the
"A" containment sump pump, indicating a leak in containment.
Initial estimates of the magnitude of the leak were approximately 300 gallons per day.
Subsequent chemistry samples indicated that there was a
small secondary leak from either a steam or feedwater line.
On September 25, 1989, two containment entries were made in order to locate and isolate the leak.
The first entry identified the general location near the manway entry area for the "A" steam generator (S/G).
The entry was terminated because further entry into the S/G area was not authorized by the radiation work permit; maximum radi--
ation levels were approximately 4 R/hr.
The second entry was made
under a special work permit and pinpointed 'the leak at a drain line valve in the "A" S/G blowdown sample line.
Control room personnel were informed and directed the operator making the entry to isolate the sample line both upstream and downstream of the leaking valve.
The radiation dose received by the person performing these isolations was 280 mrem; the health physics technician providing supervisory coverage received 250 mrem.
A maintenance work request was initiated after the leak was isolated and a work package was prepared.
On September 26, 1989, a pipe fitter was sent into containment to cap the valve.
This task was accomplished successfully.
The dose re-ceived from this activity was approximately 450 mrem to the pipe fitter and 300 mrem to the health physics technician.
This blowdown sample line was left isolated until October 4,
1989, when an operator went into containment to realign the isolated valves.
During the interim period, an alternate valve alignment was placed into service to take S/G blowdown line chemistry samples.
On October 4,
1989, an increase in the activity level from radiation monitor R-12, "Containment Gas",
was identified by the control room operations staff.
Normal R-12 readings were in the range of 1400 counts per minute (cpm); levels had increased to approximately 2600 cpm.
On October 6, 1989, with no discernible increase in primary leakage, a containment entry was made by an operations staff member (escorted by a HP technician)
in an attempt to locate a leak.
No indications of a leak were found.
On October 6, 1989, with the R-12 level still increasing and near the alarm setpoint of 15,000 cpm, a Plant Onsite Review Committee (PORC)
meeting was held to discuss the effects on plant safety of increasing the R-12 alarm setpoint to 25,000 cpm.
No adverse effects on plant safety were identified and the alarm setpoint was increased.
On October 9, 1989, a plant staff meeting was held to discuss the options available to resolve the problem of the steadily increasing R-12 activity.
At this time, the R-12 radiation level had increased to approximately 17,700 cpm.
Based on the calculated primary leak rate, primary system activity levels, and the rate at which R-12 was increasing, the staff agreed that the most probable location for the leak was from the pressurizer vapor space area.
A containment entry
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was made and a leak was identified coming from the packing gland of the "A" pressurizer spray valve.
This appeared to be the only valve with indication of a leak.
The packing gland was cleaned of boric acid deposits and then tightened to stop the leakage.
Also 'initiated on this date was a containment ventilation minipurge.
This action was taken to reduce the activity level of the containment gas, which had increased to approximately 19,700 cpm.
All required radiological permits needed for this continuous release'o atmosphere were obtained from Health Physics
& Chemistry.
Site boundary dose
from this, release was calculated at less than 1 percent of the allow-able Technical Specification limit. Activity levels in conta'inment began to drop and, by October 10, 1989, had decreased to 1'ess than 8,000 cpm.
The minipurge was stopped at approximately 12:00 p.m.
that afternoon, when R-12 was reading about 5,200 cpm.
By 4:00 p.m.,
however, R-12 was back up to approximately 6,000 cpm, indicating the suspected spray valve leak may not have been the cause of the in-creasing containment gas activity.
The inspectors will continue to follow this item until resolution is achieved.
No unacceptable licensee actions were identified during these reviews.
b.
Radiolo ical Controls (71707)
The resident inspectors periodically verified that RWPs were imple-mented properly, dosimetry was correctly worn in controlled areas, dosimeter readings were recorded, access control at entrances to high radiation areas was adequate, personnel used contamina.ion monitors as required when exiting controlled areas, and postings and labeling were in compliance with regulations and procedures.
On October 4, 1989, in order to witness a surveillance activity, a
containment'- entry was made by the inspector.
The entry team con-sisted of two maintenance workers, a quality control inspector, and a
health physics technician.
All applicable procedural requirements were verified as having been met and good ALARA practices were wit-nessed.
Maintenance (62703)
The inspectors observed portions of various safety-related mainten-ance activities to verify that redundant components were operable, activities did not violate limiting conditions for operation, re-quired administrative approvals and tagouts were obtained prior to initiating work, approved procedures were used or the activity was within the "skills of the trade," appropriate radiological controls were implemented, ignition/fire prevention controls were properly implemented, and equipment was properly tested prior to returning it to service.
Portions of the following maintenance activities were observed:
Maintenance procedure (M)-11.4.8, "Charging Pump Internal Valve Inspection and'eplacement Unit ¹18," revision 6, effective 4/16/89, observed 9/20/89...
M-37. 130,
"Disassembly 5 Reassembly of Pipe Flange Connections,"
revision 3, effective 7/5/89, observed 9/20/89.
M-37. 134,
"Replace Section of QA Welded Pipe/Equipment," revi-sion 1, effective 7/5/89, observed 9/20/8 M-37.133, "Coupling and Uncoupling "Swagelock" Tube Fittings,"
revision 0, effective 4/30/89, observed 9/26/89, Station Modification procedure (SM)-4933. 1, "PT-478, PT-479, and PT-483 (S/G "B") Tubing Reroute Upgrade," revision 0, effective 8/24/89, observed 9/25/89 and 9/26/89.
Procedure Change Notice (PCN)-89T-1581,
"Rework and Testing of PT-483," effective 9/26/89, observed 9/27/89.
All maintenance activities witnessed were performed in accordance with approved procedures by qualified workers.
With regard to the charging pump maintenance activity, a conscientious effort by Health Physics (HP) technicians responsible for overseeing the work resulted in the minimization of dose received by the maintenance workers.
In addition, HP's initial cleansing of. the area resulted in a signi-ficant reduction in the potential contamination which could have af-fected the workers during their activities.
Management oversight for this work activity demonstrated both good planning and good ALARA addressal.
The rework and testing of PT-483 is noteworthy.
A contract laborer inadvertently nicked a piece of recently installed pressure tubing with a power grinder while removing remnants of an old pipe support nearby.
The prompt notification of this incident to the licensee was an instance of good safety performance by a contract employee.
d.
Surveillance (61726)
The inspectors observed portions of survei llances to verify that test instrumentation was properly calibrated, approved procedures were used, work was performed by qualified personnel, limiting conditions for operation were met, and systems were correctly restored following testing.
Portions of the following surveillance activities were ob-served:
Refueling Shutdown Surveillance Procedure (RSSP)-5,0;
"Immediate Boration System," revision 10, effective 12/28/88, observed 9/15/89.
Periodic Test (PT)-17.2,
"Process Radiation Monitors R-11.- 22 Iodine Monitors R-10A and R-10B," revision 79, effective 9/12/89, observed 9/20/89.
PT-12. 1,
"Emergency Diesel Generator lA," revision 49, effective 8/15/89, observed 9/20/89.
PT-2.5,
"AOV's, quarterly Surveillance Auxiliary Building," re-vision 27, effective 1/17/89, observed 9/25/8 M-40, "Surveillance and Maintenance of Hydraulic Snubbers,"
re-vision 30, effective 9/29/89, observed 10/4/89.
Calibration Procedure (CP)-214, "Calibration and Maintenance of RMS Channel R-14 (Plant Vent Gas)," revision 7, effective 8/29/88, observed 10/5/89.
CP-44, "Calibration and/or Maintenance of Power Ranger N-44,"
revision 15, effective 1/30/89, observed 10/9/89.
All of the above surveillance tests were completed satisfactorily and
.in accordance with their applicable procedure.
PT-12. 1 resulted in an unanticipated D/G overspeed trip (as discussed in section 3a), but subsequent retest provided adequate assurance that the D/G was oper-able.
Good judgement by the operators was utilized in determining the impact of the overspeed trip on future D/G operability.
CP-214 found R-14 out of calibration by -22 percent in the conservative direction and it was adjusted to within required specifications.
Plant management is evaluating increasing the frequency of performing CP-214.
'~Secucit (71707)
During this inspection period, the resident inspectors verified whether x-ray machines and metal and explosive-detectors were opera-tional, Protected Area (PA) and Vital Area (VA) barriers were well maintained, access control during security turnover was adequate, personnel were properly badged for escorted or unescorted access, and compensatory measures were implemented when necessary.
On September 23 and 24, 1989, Ginna Station held its first Open House for plant workers, contractors, and their families.
An inspector attended the Open House and found'lant operations were unaffected by the large number of visitors.
Adequate planning appeared to ensure both plant security and visitor safety were maintained throughout the weekend.
With regard to the latter, posting and/or marking of the turbine building tour paths highlighted items of interest for the visitors and warned them of the safety hazards associated with the equipment.
The security staff was well prepared for the Open House as demonstrated by their increased numbers and their familiarity with
'visitor escort procedures.
Areas made accessible to visitors in-cluded the the turbine building, discharge canal area, the admini-stration building, and the simulator.
While the control room was off limits to all visitors, simulator use did allow visitors to witness normal and emergency plant evolutions.
Periodic and S ecial Re orts (90713)
Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specifications 6.9. 1 and 6.9.3 were reviewed.
This review included the following considerations:
reports contained
e
information required by the NRC; test results and/or supporting in-formation were consistent with design predictions and performance specifications; and reported information was valid.
The following report was reviewed:
Monthly Operating Report for August, 1989.
No unacceptable conditions were identified.
Evaluation of Licensee Self-Assessment Ca abilities (40500)
On September 26, 1989, a Nuclear Safety Audit and Review Board (NSARB) meeting was held at Ginna Stations The minimum number of board members required per Technical Specification 6.5.2 was verified present.
The inspectors noted that the meeting was well organized and that the discussion among board members was open, honest, and constructive.
The board addressed all NRC inspection report viola-tions and tho licensee's corrective action plans for their closeout.
Additional attention was focused on the plant's most recent Systema-tic Assessment of Licensee Performance (SALP) ratings and plant man-agement's as;essment of these ratings.
The discussion centered around plant management's continued effort to coordinate an improved self-assessment program.
Specifically, board members expressed con-cern that the Quality Assurance/Control (QA/QC) organization was not being effectively -utilized in identifying performance trends to plant management.
The NSARB chairman stated that the underutilization of QA/QC was currently being looked at, and that appropriate actions needed to address this concern would be taken.
The inspectors concluded the NSARB had provided adequate independent review and audit of activities in plant operations, chemistry and radiological safety, instrumentation and control, and quality assur-ance practices.
Fire Bri ade Shift Mannin (71707)
Ouring the Emergency Operating Procedure (EOP) inspection conducted between September 25, 1989, and October 4, 1989 (see Inspection Re-port 50-244/89-80),
a concern was brought up by the team leader that the Ginna Station Fire Brigade shift manning complement did not meet the requirements of Appendix R, section III. H.
Specifically, Fire Brigade Captains were neither in possession of an operator's license nor had the equivalent training in plant safety-related systems to understand effects of fire and fire suppression components on safe shutdown capability.
After documentation review by the inspector, RGEE corporate licens-ing, and the NRC Office of Nuclear Reactor Regulation, it was deter-mined Ginna Station was not required to be committed to section III.
H. of Appendix R.
Through a letter dated January 23, 1980, the NRC
'
amended RG&E's license "to maintain a Fire Brigade of 5 members."
The submittal letter for this license amendment request, dated November 11, 1979, stated that all five members of the Fire Brigade would be equally trained with at least two of these members having additional leadership training for the Brigade Captain position.
In reviewing this commitment, the licensee found that not all shift Fire Brigade crews had two qualified Brigade Captains.
The Opera-tions Manager's actions to remedy this situation. consisted of the following: informing all Shift Supervisors that all Fire Brigades must have at least two qualified Brigade Captains;,ensuring that scheduling of personnel onto shifts would allow for the assignment of two Brigade Captains; and, for those individuals who were identified as not being Brigade Captain qualified, ensuring that the requisite training had been completed and that qualification letters were issued.
The inspectors verified these actions to be acceptably com-pleted before the end of this inspection period.
The inspector had no further concerns.
4.
Exit Interview (30703)
The inspectors met with senior plant management periodically and at the end of the inspection period to discuss inspection scope and findings.
Based on NRC Region I review of this report and discussions held with licensee representatives, it was concluded this report does not contain information subject to
CFR 2.790 restriction '