IR 05000244/1991010
| ML17262A515 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/31/1991 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17262A514 | List: |
| References | |
| 50-244-91-10, NUDOCS 9106180082 | |
| Download: ML17262A515 (15) | |
Text
U. S. NUCL Inspection Report 50-244/91-10 License: DPR-18 EAR REGULATORY COMMISSION
REGION I
Licensee:
Facility:
Inspection:
Inspectors:
Approved by:
Rochester Gas and Electric Corporation (RG&E)
R. E. Ginna Nuclear Power Plant April 15 through May 20, 1991 T. A. Moslak, Senior Resident Inspector, Ginna N. S. Perry, Resident Inspector, Ginna P. P..Sena, Reactor Engineer, PB3, DRP E. C. McCabe, Chief, Reactor Projects Section 3B INSPECTION SCOPE s l~]l~l Date Plant operations, radiological controls, maintenance/survdeillance, emergency preparedness, security, engineering/technical support, and safety assessment/quality verification.
INSPECTION OVE<RVIEW Adddy f fig g
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p rigidly.controlled.
The new Advanced Digital Feedwater Control System, a major modification, performed as designed during power escalation.
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<50%; operators manually stabilized the reactor at 2% power using condenser steam dumps and the atmospheric relief valves.
Under the control of the turbine's auxiliary governor, the 20-AG-2 solenoid valve normally" provides a means of preventing turbine overspeed in the event of a complete load loss.
As a result of the failed 0-ring seal, the Electro-Hydraulic (EHC) fluid was dumped from the control and intercept valves to the EHC drain.
The closing of these valves caused a reverse power condition after the entrapped steam had been used and resulted in the generator output breakers opening and the turbine stop valves closing.
Maintenance personnel conducted an in-depth investigation into the cause of the 0-ring failure on 20-AG-2.
The failure was attributed to the replacement of the upper valve block with a block that did not seat exactly on the lower block.
Through machining the mating surface of the upper block, the problem was corrected.
The inspectors reviewed the maintenance department follow-up actions to this incident and concluded that a thorough evaluation was performed in a timely manne.2 Surveillance Observations Inspectors observed portions of surveillances to verify proper calibration of test instrumentation, use of approved procedures, performance of work by qualified personnel, conformance to Limiting Conditions for Operation (LCOs), and correct system restoration following testing.
The following surveillances were observed:
Refueling Shutdown Surveillance Procedure (RSSP)-2.1, Safety Injection Functional Test, Revision 40, effective April26, 1991, observed April 29, 1991 RSSP-2.2, Diesel Generator Load and Safeguard Sequence Test, Revision 39, effective April 2, 1991, observed April 30, 1991.
The inspectors observed good management oversight during performance of these surveillances and concluded that overall coordination and control of the surveillances was good.
No inadequacies were identified.
4.0 EMERGENCY PREPAREDNESS 4.1 Emergency Preparedness Coordination Meeting On May 14, 1991, the inspectors attended a joint meeting of the RG&E Emergency Preparedness staff, FEMA, and members of the State and County Emergency Response organizations.
The purpose of the meeting was for the participating organizations to further define the extent of their involvement in the full participation annual exercise scheduled for September 11, 1991.
5.0 SECURITY 5.1 Routine Observations During this inspection period, the resident inspectors verified that x-ray machines and metal and explosive detectors were operable, protected area and vital area barriers were well maintained, personnel were properly badged for unescorted or escorted access, and compensatory measures were implemented when necessary.
No unacceptable conditions were identified.
5.2 Revised Security Plan On May 1, 1991, the RG&E Security Department submitted a revised Security Plan to the NRC for approval.
On May 22, 1991, RG&E representatives met with NRC security staff in the regional office to discuss elements of the revised pla.0 ENGINEERING/TECHNICALSUPPORT 6,1 Elongated Steam Generator Manway Stud On April 21, 1991, the manway was torqued onto the "A" Steam Generator cold leg.
The Number 8 Stud exceeded the acceptance criterion for elongation.
A technical justification, prepared by RG&E Corporate Engineering on April 25, allowed plant operation to cold
'shutdown conditions of 200 degrees maximum and 410 psig maximum.
Through observations and measurements, RG&E concluded that the top of the stud was bent approximately 1 degree.
It was not possible to determine ifproper stud preload and elongation of the load-carrying portion of the stud had been obtained, so a Nonconformance Report (NCR) was initiated.
The NCR was dispositioned as acceptable for Use-As-Is based on the following considerations.
The probable elongation of the stud did not occur in the load-carrying portion of the stud, and all ASME Section III Code requirements are met even ifthe stud is assumed to be failed.
RG&E engineering concluded that the most probable cause of the measured elongation was plastic deformation of the reduced diameter upper portion of the stud, which is not load-carrying.
This deformation was due to an angular misalignment between. the piston and reaction nut on the hydraulic tensioner, resulting from the piston hitting the tensioner mechanical stops eccentrically during initial tensioning.
Further analysis showed that with conservative assumptions the lower portion of the stud, which is load-carrying, is probably within the required range of elongation.
However, RG&E contracted an engineering firm to perform an analysis for pre-load, design, and heat-up conditions in the primary manway, assuming the stud has failed.
Based on this analysis, RG&E concluded that the effect of the hypothetical failed stud is minor.
With the stud failed, all ASME Code allowable requirements are still met under design, normal, operating, and heat-up transient conditions.
Additionally, Engineering recommended that the stud be replaced during the 1992 outage, and that the tensioning procedure, be revised to prevent future occurrences of piston/reaction nut misalignment.
The inspectors reviewed the NCR, the engineering analysis assuming a failed stud, and the 10 CFR 50.59 Safety Evaluation.
The analysis and Safety Evaluation were detailed and comprehensive in addressing the elongation of the steam generator manway stud, and the NCR was properly reviewed and dispositioned.
Good management concern and involvement were observed during resolution of the problem.
7.0 SAFETY ASSESSMENT/QUALITY VERIFICATION 7.1 Periodic Reports Periodic reports submitted by the licensee pursuant to Technical Specifications 6.9.1 and 6.9.3 were reviewed.
Inspectors verified that the reports contained information required by the NRC, that test results and/or supporting information were consistent with design predictions and performance specifications, and that reported information was accurate.
The following reports were reviewed:
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I e
Monthly Operating Reports for March and April 1991.
No unacceptable conditions were identified.
7.2 Licensee Event Reports (LERs)
LERs submitted to the NRC were reviewed to determine whether details were clearly reported, causes were properly identified, and corrective actions were appropriate.
The inspectors also assessed whether potential safety consequences were properly evaluated, generic implications were indicated, events warranted onsite follow-up, and applicable requirements of 10 CFR 72 and 10 CFR 73 were met.
The following LERs were reviewed (Note:
date indicated is event date):
91-003, Firewatch Not Posted During Fire System Isolation, March 14, 1991.91-004, Manual Start and Loading of "B" Emergency Diesel Generator Due to Potential Severe Weather Condition, March 28, 1991.91-005, Steam Generator Tube Degradation Due to IGA/SCC Causes Q.A. Manual Reportable Limits to be Reached, April 14, 1991.
The events associated with the above LERs were reviewed in NRC Inspection report 50-244/91-07.
The inspectors concluded that the LERs were accurate and met regulatory requirements.
No unacceptable conditions were identified.
A 8.0 ADMINISTRATIVE 8.1 Inspection Hours This inspection included 5 backshift and 19.5 deep backshift hours.
8.2 Exit meetings At periodic intervals and at the conclusion of the inspection, meetings were held with senior station management to discuss the scope and findings of this inspection.
In addition, an NRC exit meeting was held for 50-244/91-11 (SSFI/MTI follow-up) on April 26, 1991.
The exit meeting for inspection report 50-244/91-10 was held on May 22, 1991 with the following individuals attending:
~Nm Clair Edgar Duane Filkins Paul Gorski Alan Jones Thomas Marlow Thomas Moslak Robert Wilkinson Joseph Widay
~Titl Mgr. Electrical/I&C Mgr, HP & Chemistry Mgr. Mech. Maintenance Corrective Action Coordinator Superintendent, Support Services SRI-NRC QC Eng.-Main.(for Mike Lilley, Nuc. Assurance Mgr.)
Superintendent, Ginna Production 8.3 Information Meeting with Local Officials On May 16, 1991, the inspector met with the local Town Supervisor, at the supervisor's office,'in Ontario, New York. During this informal meeting, the responsibilities of the NRC and the role of the resident inspector at the'Ginna site were discussed.
Current plant status and general topics of interest were also discussed.
A follow-up meeting to introduce additional key NRC personnel associated with the facility willbe scheduled at a later date.