IR 05000244/1988005
| ML17251B070 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/27/1988 |
| From: | Cowgill C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17251B068 | List: |
| References | |
| 50-244-88-05, 50-244-88-5, IEB-80-03, IEB-80-3, NUDOCS 8805090263 | |
| Download: ML17251B070 (20) | |
Text
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U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
Docket No.
50-244/88-05 50-244 Licensee No.
Licensee:
DPR-18 Priority Rochester Gas and Electric Corporation 49 East Avenue Rochester, New York Category C
Facil'ity:
Location:
Dates:
Inspectors:
Approved by R.
E. Ginna Nuclear Power Plant Ontario, New York March 7 - Apri 1 5, 1988 C.
S. Marschall, Senior Resident Inspector, Ginna N.
S. Perry, Resident Inspector, Ginna C. J.
Cowg'
Chief, Reactor Projects Section 1A
>7~
g'5'ate Ins ection Summar A~I:R
- i b
h ties including followup of open items, plant startup after refueling, plant shut-down for a primary to secondary leak repair, plant startup after leak repair, safety system walkdowns, operational safety verification, surveillance, maintenance, physical security, onsite followup of events, review of written reports, Loss of Charcoal from Standard Type II, 2 inch, Tray Adsorber Cells.
This inspection in-volved 193 hours0.00223 days <br />0.0536 hours <br />3.191138e-4 weeks <br />7.34365e-5 months <br /> by the inspectors, including 39 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> of backshift inspection coverage.
Results:
A weakness identified in control of activities affecting quality is dis-cussed in Section 8.
One violation resulted from this.
The need for increased attention to housekeeping is discussed in Section 5.
SS05090263 SS0503 PDR AOOCK 05000244 Q
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:
J.
C. Bodine, Nuclear Assurance Manager
"E.
C.
Edgar, Instrument and Control Supervisor D.
L. Filkins, Chemistry
& Health Physics Manager
"P. J. Gorski, Station Engineer
"J.
L. Hotchkiss, Modification Project Manager R.
la'.
Kober, Senior Vice President, Production and Engineering R. A. Marchionda, Training Manager T. A. Marlow, Maintenance Manager T. A. Meyer, Superintendent Ginna Support Services
- J.
B. Neis, Station Engineer
~J.
T. St. Martin, Station Engineer T.
R. Schuler, Operations Manager
"L. F. Smith, Operations Supervisor B. A. Snow, Superintendent Nuclear Production
- S.
M. Spector, Superintendent Ginna Station
~S.
B. Warren, Health Physicist
~J.
A. Widay, Technical Manager
~R.
E.
Mood, Supervisor Nuclear Security
"Denotes persons present at Exit Meeting on April 6, 1988.
Followu on Previousl Identified Items Closed Ins ector Followu Item 87-04-02:
Resolution and Corrective Action of Diesel Fuel Oil Strainer Clogging.
This item addressed the licensee's long term corrective actions associated with diesel fuel oil strainer clogging.
The fuel oil storage tank cleaning procedure was revised to eliminate fibrous contaminants.
Monthly Emergency Diesel Generator surveillance tests were revised to include new data sheets to monitor the fuel oil transfer system.
The inspectors reviewed installation of diesel fuel oil duplex strainers, including the following procedures:
SM-4526.7, Diesel Fuel Oil Duplex Strainer Installation Mechanical, Revision 0.
M-15. 1, A or B Diesel Generator Inspection and Maintenance, Revision
PT-12. 1, Emergency Diesel Generator IA, Revision
'
PT-12.2, Emergency Diesel Generator 1B, Revision
The inspectors found the licensee's corrective actions to control strainer clogging adequate and implemented in a timely manner.
Closed Ins ector Followu Item 87-18-01
- Failure to Recognize and Ac-knowledge Technical Specification TS) Limiting Conditions for Operation (LCOs).
This item identified a concern with the licensee's timeliness in entering an LCO after receiving out of specification diesel fuel oil sample test results.
In a response dated October 9, 1987, the licensee stated its position is "to assure strict compliance to TS requirements in the event of unacceptable results in any surveillance".
When successive samples differ, significant procedural changes were made, or factors existed which could have affected test results, resampling and reanalysis are required to determine the validity of preliminary results.
The inspectors found the licensee's stated policy was reasonable in addressing potential concerns of test data validity.
Licensee personnel are sensitive to obtaining confirming data in a timely manner.
The inspectors had no fur-ther questions.
Review of Plant 0 erations The plant was in cold shutdown making preparations for startup at the begin-ning of the inspection period.
Criticality was achieved March 9, 1988.
On March 10, 1988, the reactor tripped automatically from 25% power due to low steam generator level combined with a steam flow feed flow mismatch.
Operator error caused the trip.
On March ll, 1988, two main steam safety valves opened due to high main steamline pressure.
Operators were inexperienced with a higher than usual positive Moderator Temperature Coefficient (MTC).
After retraining operators using the site specific simulator, the turbine generator was successfully synchronized to the grid on March 12, 1988.
On March 14, 1988, operators shutdown the plant from 89% power to repair a
tube leak in the B steam generator.
An Unusual Event was declared and was terminated about 4 1/2 hours later.
A total of nine tubes were plugged in the B steam generator, discussed in section 13.
A problem identified after tensioning the B steam generator primary manway covers, discussed in Section 8, resulted in one violation.
The reactor was taken critical and the main generator was synchronized to the grid on March 24, 1988.
At the close of the inspection period, the plant was operating at approximately 100% power.
En ineered Safet Feature ESF S stem Walkdown A complete walkdown of accessible portions of the Auxiliary Feedwater (AFW)
system was performed to verify its operability.
The'nspectors verified the licensee's lineup matched plant drawings and as'built configurations.
System components were checked for labeling, cleanliness, and state of repair".
In
three instances, local instrument isolation valve positions differed from as-built drawings, and two valves were missing labels.
The inspector provided the discrepancies to licensee management.
No conditions adverse to safety were identified.
0 erational Safet Verification On a daily basis, inspectors observed shift turnover and conduct of operations-in the control room.
Operator performance is discussed in Section 3.
During the previous inspection period, housekeeping varied considerably, and warranted greater licensee attention (section 6 of IR 50-244/88-01).
During the present inspection period, frequent tours by resident inspectors and a
tour by NRC Region I management personnel indicated increased management at-tention is still warranted in most areas, with notable exception of inside containment.
The licensee's program to assure quality of housekeeping prior to plant startup was effective in containment; response to resolution of this issue, especially in the Auxiliary and Intermediate Buildings, has been slow.
Many areas contain temporarily stored items; some are massive, unrestrained, and near or in contact with safety-related equipment.
Areas not usually
"dirty", contain items not properly stored, including spare light bulbs, extra strip chart paper, cables and procedures.
Labelling and identification of plant equipment is inconsistent (see paragraph 4).
Plant management is ad-dressing these areas, and expects housekeeping to improve dramatically during the first few weeks of April.
Licensee assurance of quality in housekeeping needs increased management at-tention.
Resolution of safety implications of unrestrained objects near safety-related equipment needs to receive more timely action.
Monthl Surveillance Observation Inspectors observed portions of surveillance test procedures listed below to verify test instrumentation was properly calibrated, approved procedures were used, work was performed by qualified personnel, Limiting Conditions for Operations were met, and the system was correctly restored following testing.
The following surveillance activities were observed:
PT-16, Revision 48, "Auxiliary Feedwater System", effective date February 2,
1988, observed March 10, 1988.
Licensee controls and procedures were adequate to insure surveillance activi-ties were conducted in accordance with license requirements.
Monthl Maintenance Observations The inspectors observed portions of various safety-related maintenance acti-vities to determine redundant components were operable, activities did not violate Limiting Conditions for Operation, required 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 radio-
logical controls were implemented, ignition/fire prevention controls were properly implemented, and equipment was properly tested prior to returning it to service.
Calibration CP-216, "Calibration and/or Maintenance of RMS Channel R-16",
effective date January 8,
1986, observed April 4, 1988.
Licensee control of this activity was considered adequate.
Maintenance (M)-43. 15.2, "Installation of B Steam Generator Cold/Hot Leg Primary Manway Cover Utilizing Hydraulic Stud Tensioner.
Licensee activities related to this maintenance observation are discussed in Section 8.
In Inspection Report 50-244/88-01, inspectors noted that, during a Residual Heat Removal pump post-maintenance test, the pump was stopped before oil levels had stabilized.
The licensee is making appropriate procedural changes to applicable test procedures regarding pump oil level measurements to insure consistent monitoring during test performance.
An Inter-office Correspondence was distributed addressing pump oil level monitoring during testing; accept-ance criteria of stable level, not less than 1/3 full (cup level),
was estab-lished.
The licensee's actions addressed the concern in a timely manner and exhibited appropriate level of management involvement and control in assuring quality when returning safety-related equipment to service following maintenance'team Generator Manwa Cover Stud Tensionin On March 20, 1988, after conclusion of tube plugging in the B Steam Generator, maintenance personnel performed maintenance procedure M-43: 15.2, revision 3, effective date February 24, 1988, "Installation of B Steam Generator Hot/Cold Leg Primary Manway cover Utilizing Hydraulic Stud Tensioner"
~
As required by step 5.21.23 of the procedure, Quality Control (QC) inspectors took meas-urements to determine if stud elongation was within the acceptance criterion of.0075
+.0010 inches.
Each manway had several studs which did not meet this acceptance criterion and the maintenance personnel were informed.
The.
procedure for each manway cover installation was completed and signed by the mechanical maintenance personnel and QC inspectors.
Non-Conformance Reports (NCR) were written by the QC inspectors requiring corrective action for un-acceptable stud elongations prior to startup.
Neither maintenance nor QC informed operations personnel or site management of the unacceptable stud elongations or of the existence of the NCRs.
Operations personnel filled and vented the Reactor Coolant System (RCS),
then began to heat up in preparation for an RCS hydrostatic test.
On March 21, 1988, a Quality Control Engineer discovered the NCRs and notified the station superintendent that the accept-ance criteria for tensioning both B steam generator primary manway covers had not been met.
The licensee depressurized and drained the RCS and successfully repeated the tensioning procedure for both B steam generator manway cover During review of the event, inspectors identified that step 5.21.23 of proce-dure M-43. 15.2 requires previous steps be repeated for stud tensioning until acceptable stud elongation is achieved.
This ensures adequate corrective action when acceptance criteria for stud elongation are not met.
The mechani-cal maintenance personnel and gC inspectors did not sign step 5.21.23; however, they did sign subsequent steps indicating completion of the procedure.
Technical Specification 6.8. 1 requires that the licensee establish and imple-ment procedure.a The failure to follow the provision of TS 6.8. 1 and Proce-dure M-43. 15.2 described above is an apparent violation (VIO 50-244/88-05-01).
The guality Assurance Manual, Ginna Station, which applies to activities af-fecting quality, requires, in part:
the responsible supervisor be informed when quality related activities are not in accordance with documented instructions, verification acceptance be completed and released prior to continuing a quality activity, and quality control inspection be performed by individuals other than those who performed the maintenance activity.
Site gC and maintenance personnel failed to meet these requirements.
Based on interviews with site management personnel, the inspectors identified a lack of cooperation between departments in control. of activities affecting quality.
Various department managers do not respect gC personnel, and gC management does not have confidence that other department personnel have an appropriate concern for quality control.
The lack of cooperation is known by site superintendents and corporate management and is indicative of highly motivated people involved in interdepartmental disputes.
This resulted in a fai lure to effectively communicate safety concerns to proper levels of man-agement for resolution.
This is considered a weakness in the licensee's pro-gram for assurance of quality.
9.
Ph sical Securit Review The inspectors made observations to verify selected aspects of the station physical security program were in accordance with regulatory requirements, including the physical security plan and approved procedures.
During this inspection period, observations and interviews were conducted to verify lic-ensee control of entry and exit for packages and material entering the pro-tected area was in conformance with the physical security plan and regulatory requirements.
Licensee procedures were adequate to control packages and material entering and exiting the protected are Review of General Em lo ee Trainin GET Classroom training on the purpose and use of respiratory protection equipment was reviewed to evaluate the effectiveness of training for non-licensed plant'taff, contractors, and visitors having responsibilities in the area of radi-ation safety.
Presentations included lectures, discussion, video tapes and a practical demonstration of donning a respirator with class participation.
Presentations were accurate, complete, and professionally presented.
The licensee's program for GET in respiratory protection was adequate to in-sure proper training for radiation workers.
Onsite Followu of Plant Events On March 10, 1988 at 6:56 P.N., the reactor tripped automatically from 25%
power due to a low steam generator level combined with a steam flow feed mis-match.
Operator error, due to inexperience with a higher than usual positive moderator temperature coefficient (MTC), was determined to be the cause of the trip.
On March 11, 1988, the onsite simulator was re-programmed to model the positive MTC and operators practiced rolling the turbine and synchronizing
.the main generator under the current plant conditions on the simulator.
On March 12, 1988, and again on March 13, 1988, after turbine overspeed testing, operators successfully rolled the turbine and synchronized the main generator without fur ther difficultie s.
On March 14, 1988 at approximately 3:00 P.M.,
when the Steam Jet Air Ejector radiation monitor alarmed, operators initiated a request for chemical sampling of the air ejectors and steam generator blowdown system.
Analysis of samples conflicted; one result was above Technical Specification (TS) limits of 0. 1 gpm leakage into the B steam generator, and the other below.
At 6:45 P.M.,
the licensee decided to begin a shutdown from 89% power while new samples were being analyzed.
All rods were fully inserted and the reactor trip breakers manually opened at 9:27 P.M.
When new blowdown sample results, received at 9:30 P.M., confirmed leakage of 0. 136 gpm into the B steam generator, an Un-usual Event was declared.
The Unusual Event was terminated at 2: 13 A.N ~
on March 15, 1988, when RCS temperature was less than 350 degrees Fahrenheit, and the TS limit on leakage no longer applied.
See paragraph 13 for ensuing activities.
No violations were identified.
Plant Startu From Refuelin At the end of the 1988 refueling outage the inspectors verified systems dis-turbed during the outage were returned to an operable status prior to startup, and plant.startup, heatup and approach to critically were conducted according to approved procedure A walkthrough to determine systems disturbed during the outage were properly returned to service was performed for portions of the following systems:
Emergency Diesel Generators Source Range Detectors Safety Injection Containment Spray Feedwater In addition, startup activities were monitored to insure compliance with technically adequate, current, and approved procedures, requirements of tech-nical specifications were met, and surveillance tests were satisfactorily performed when required.
The licensee used a controlled-approach prior to and during the startup.
Daily meetings were effectively used to address concerns and keep plant man-agement informed about plant status.
Requirements and concerns were dealt with in a timely, conservative manner.
No unacceptable conditions were identified.
Forced Outa e Steam Generator Tube Plu in After the plant was shutdown on Narch 14, 1988, personnel looked for addi-tional tubes to plug by: performing a secondary hydrostatic test and visually observing leaks exhibiting greater than one drop per minute; reviewing Eddy Current Test results from the refueling outage; and performing additional Eddy Current Testing.
Two tubes were plugged as a result of leaks found dur-ing hydrostatic testing, five tubes from re-evaluation of refueling outage Eddy Current Test results, and two tubes from results obtained during addi-tional Eddy Current Testing.
Of the nine tubes plugged, four were attributed to personnel errors in reviewing Eddy Current Test results obtained during the refueling outage.
During the 1988 refueling outage, initial data analysis review was performed by ZETEC's Computer Data Screening (CDS) System, and data analysis was performed utilizing the ZETEC DDA-4 Digital Data Analysis System.
Potential steam generator tube problems identified by computer were resolved by an RG&E reviewer.
In the future, RG8E plans to have discrepancies between computer and RG5E reviewer evaluated by an additional RGEE reviewer.
The licensee's actions were considered conservative in dealing with the steam generator tube plugging.
Senior site and corporate management attention and involvement were evident during the outage.
As data became available during additional steam generator'tube testing, upper management responded in a timely manner while maintaining conservatism approach to operational safety.
Communication between the licensee and the NRC during the forced outage was good; the NRC was informed of test results and licensee decisions in a timely manner.
One problem was identified with tensioning of the steam generator manway (see paragraph 8).
No other conditions adverse to safety were identifie '
14.
Loss of Charcoal - IE Bulletin No. 80-03 The inspectors reviewed the licensee's response to NRC IE Bulletin No. 80-03, dated March 21, 1980, for accuracy and completeness.
In the response, RG&E stated two repairs needed to be made; B Containment Purge Filters and A and B Auxiliary Building Charcoal Filters needed to be changed.
The inspectors verified both had been changed.
Additionally, the inspectors reviewed pro-cedure PT-38. 1, "Visual Inspection of Charcoal Filter Tray Assemblies",
re-vision 5, effective date November 2, 1984, and found the procedure,to ade-quately address IE Bulletin No. 80-03 concerns.
No violations were identified during the review.
This Bulletin is considered closed.
15.
Review of Written Re orts of Nonroutine Events Written reports submitted to the NRC were reviewed to determine whether de-tails were clearly reported, causes properly identified and corrective actions appropriate.
The inspectors also determined whether assessment of potential safety consequences'had been properly evaluated, generic implications were indicated, events warranted onsite follow-up, reporting requirements of 10 CFR 50.72 were applicable, and requirements of 10 CFR 50.73 had been properly met.
The following LERs were reviewed and found to be satisfactory (Note: dates indicated are event dates):
88-001, 3/7/88, Higher than normal count rate on source range NIS due to faulty connectors, causes reactor trip during source range re-ener-gization.
The following 10 CFR 21 reports were reviewed and found to be satisfactory (Note: dates indicated are reporting dates):
March 31, 1988, Amptector Actuator Arm Clearance.
No unacceptable conditions were identified.
16.
Review of Periodic and S ecial Re orts Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specifications 6.9. 1 and 6.9.3 were reviewed by the inspectors.
This review included the following considerations:
reports contained informa-tion required by the NRC; test results and/or supporting information were consistent with design predictions and performance specifications; and y e-ported information was valid.
Within this scope, the following reports were reviewed by the inspectors:
Monthly Operating Report for February 1988.
The report was considered adequate to meet regulatory requirements.
1.
~Ei M
h At periodic intervals during the inspection, meetings were held with senior facility management to discuss inspection scope and findings.
Based on.NRC Region I review of this report and discussion held with licensee representa-tives, it was determined this report does not contain information subject to
CFR 2.790 restriction ~