IR 05000244/1982010
| ML20054K904 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/15/1982 |
| From: | Kister H, Zimmerman R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20054K905 | List: |
| References | |
| 50-244-82-10, NUDOCS 8207060269 | |
| Download: ML20054K904 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION Region I Report No. 50-244/82-10 Docket No. 50-244 C
License No. DPR-18 Priority Category
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Rochester Gas & Electric Corporation Licensee:
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89 East Avenue Rochester, New York 14649 R. E. Ginna Nuclear Power Plant Facility Name:
Inspection at: Ontario, New York
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Inspection conducted: May 1-31, 1982
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Inspectors:
dat4 signed R. (ZimmeMn, Senior Resident Inspector date signed
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Approved by:
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B. Kis'te $ Chief, Reactor Projects df te s/igned HSection 1C, Division of Projects &
Resident Programs
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Inspection Summary:
1-31,1982 (Report No. 50-244/82-10)
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Inspection on May Routine, onsite, regular, backshift, and weekend inspection by the Areas Inspected:
18-20(135 hours0.00156 days <br />0.0375 hours <br />2.232143e-4 weeks <br />5.13675e-5 months <br />). Areas inspected resident inspector, and section chief on May included plant operating records; surveillance testing; maintenance; material accountability
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for steam generator modifications; licensee action on previous inspection findings; periodic -
and special reports; and accessible portions of the facility during plant tours.
Of the 7 areas inspected, one violation was identified in one area (Failure to Results:
properly implement and track QC Surveillance Reports-Paragraph 6.a).
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Region I Form 12
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(Rev. April 77)
8207c60269 820616, DRADOCK05000g
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DETAILS
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1.
Persons Contacted
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The below listed technical and supervisory level personnel were among those
contacted:
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E. Beatty, Operations Supervisor i
J. Bodine QC Engineer i
L. Boutwell, Maintenance Supervisor
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C. Edgar, I & C Supervisor l
D. Filkins Supervisor Health Physics and Chemistry l
D. Gent, Results and Test Supervisor
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G. Larizza Technical Engineer
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T. Meyer, Nuclear Engineer j
R. Morrill, Training Coordinator l
J. C. Noon, Assistant Plant Superintendent i
C. Peck, Operations Engineer i
B. Quinn, Health Physicist
B. A. Snow, Plant Superintendent S. Spector, Maintenance Engineer.
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J. Straight, Fire Protection and Safety Coordinator j
R. Wood, Supervisor of Nuclear Security
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The inspector also interviewed and talked with other licensee personnel during.
I the course of the inspection.
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l 2.
Licensee Action on Previous Inspection Findings
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(Closed) Inspector Follow Item (81-11-02): Periodic Test (PT) Procedure-2.10.3
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Low Head Safety Injection Check Valves, has been revised, requiring MOVs 852A&B
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to be maintained open for ten minutes. This allows sufficient time for back l
leakage past check valves 853A&B to be indicated by an increase on Pressure In-i dicator 639. The tast was performed on May 21, 1982 with no leakage noted.
(Closed) Unresolved Item (82-03-04): Exxon Nuclear Company, Inc. has evaluated the increased fan cooler capacity on LOCA ECCS results by recalculating the segment of the transient when the fan coolers are in operation. The analysis for nominal temperature and pressure operation was issued in Report XN-NF-82-26,
1982. The new calculated Peak Clad Temperature (PCT) is 1928 F, dated April 16, 0F over the results assuming single fan cooler operation wi h a an increase of 6 peak Linear Heat Generator Rate of 13.76 KW/ft at a total peaking factor (F ) of 2.32.
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Review of Plant Operations a.
Throughout the reporting period, the inspector reviewed plant operations associated with the annual refueling, modification, and maintenance outage and plant restart on May 24. Major activities in progress included per-formance of the Containment Integrated Leak Rate Test; repair of contain-ment purge supply and exhaust valves following leak testing; performance of start-up physics testing; and troubleshooting alarms on the 'B'
Steam Generator from the newly installed metal impact monitoring system.
b.
Shift Logs and Operating Records Operating logs and records were reviewed against Technical Specification and administrative procedure requirements.
Included in the review were:
daily during control room Control Room Log
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surveillance daily during control room Daily Surveillance Log
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surveillance daily during control room Shift Supervisor's Log
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surveillance daily during control room Plant Recorder Traces
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surveillance daily during control room Plant Process Computer Printout
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surveillance 5/1/82 through 5/31/82 Station Event Reports
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5/1/82 through 5/31/82 Maintenance Work Orders and Trouble
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The logs and records were reviewed to verify that entries were being properly made; entries involving abnormal conditions provided sufficient detail to communicate equipment status, deficiencies, corrective action restoration and testing; records were being reviewed by management; operating orders did not conflict with the Technical Specification or reporting requirements; logs and records were maintained in accordance with Technical Specification and administrative procedure requirements.
c.
Plant Tour 1.
During the course of the inspection, tours of the following areas were conducted:
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Control Room Auxiliary Building
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Intermediate Building (including control point)
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Containment
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Service Building Turbine Building
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Diesel Generator Rooms
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Battery Rooms
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Screenhouse
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Yard Area and Perimeter
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2.
The following observations resulted from the tours:
a.
Monitoring instrumentation.
Process instruments were observed for correlation between channels and for conformance with Technical Specification requirements.
b.
Annunciator alarms. Various alarm conditions which had been re-ceived and acknowledged were observed. These were discussed with shift personnel to verify that the reasons for the alarms were understood and corrective action, if required, was being taken.
c.
Shift manning. Control room and shift manning were observed for conformance with 10 CFR 50.54 (K), Technical Sp.ecifications, and administrative procedures.
d.
Radiation protection controls. Areas observed included control point operation, posting of radiation and high radiation areas, compliance with Radiation Work Permits and Special Work Permits, personnel moni-toring devices being properly worn, and personnel frisking practices.
Radiation Work Permit controls are discussed in paragraph 4.
e.
Equi 3 ment lineups. Valve and electrical breakers were verified to be in t1e position or condition required by Technical Specifications and plant lineup procedures for the applicable plant mode. This verifi-
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l cation included control board indications daily and field observations
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made during routine plant tours.
l During review of the Locked Valve List on May 14, it was noted that
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the normally locked closed, service air supply containment isolation valve, MV 7141, incorrectly entered as 7241, was logged as in the l
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open unlocked position since February 2.
Containment integrity was required on May 7 for crevice cleaning of the steam generators.
The inspector verified by review of Operating Procedure (0)-1.1B, l
Establishing Containment Integrity, that MV 7141 was in fact locked
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closed by May 7 as required, and the Locked Valve List had not been updated. The inspector discussed with licensee manage-
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ment the need to maintain the List both current and accurate. No problems have been previously noted in maintaining the Locked Valve List.
f.
Equi) ment tagging. Selected equipment, for which tagging requests had )een initiated, was observed to verify that tags were in place and the equipment in the condition specified.
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Fire protection.
Fire detection and fire fighting equipment and controls were observed for conformance with Technical Specifications and administrative procedures.
h.
Security. Areas observed for confomance with regulatory require-ments, and site security plan and administrative procedures, in-cluded vehicle and personnel access, protected and vital area integ-rity, escort and badging.
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Plant housekeeping. Plant conditions were observed for conformance with administrative procedures. Storage of material and components was observed with respect to prevention of fire and safety hazards.
Housekeeping was evaluated with respect to controlling the spread of surface and' airborne contamination.
No violations were identified.
4.
Radiation Work Permit (RWP) Cortrols.
RWP's are issued to cover work of a routine nature and are valid for an extended period, usually a calendar year.
Each person performing work under an RWP must read its requirements and sign the back of the RWP certifying that the RWP was understood and will be complied with.
In addition, whenever an individual works under an RWP that person must also document on a daily RWP Log the following in-formation: film badge number; pocket dosimeter number including dose in and out; and time in and out of the controlled area.
On May 4, the inspector noted that two individuals had signed the daily RWP Log indicating they were working under RWP 40 for the Upper Radwaste Storage Building.
Upon review it was determined the two workers had not signed the back of the RWP certifyirg the RWP was read and understood. The Health Physics Foreman recalled the workers, who indicated they had not yet entered the controlled area. Discussions with the workers verified that they were aware of all requirements pertaining to the RWP. The inspector considered the above instance to be an isolated case, not in-dicative of normal worker practice.
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RWP 40 had recently been issued on April 27, replacing RWP 15. Although the workers above were aware of the RWP change, the inspector noted that no formal method exists to inform workers when an RWP is cancelled or revised during the year. The licensee representative stated that a formal method would be developed and implemented by August 31, 1982.
5.
Reactor Coolant Pump Seal Integrity During Safety Injection Concurrent with Loss of Offsite Power During observation of functional testing on May)24, the inspector became awarePum that the 'A'
& 'B' Component Cooling Water (CCW on a Safety Injection (SI) signal coincident with undervoltage on 480 V Buses 14 and 16, respectively (loss of offsite power). Since the Charging Pumps are tripped automatically on a SI signal, seal injection to the Reactor Coolant Pumps and cooling water to the thermal barrier would be terminated upon an SI signal con-current with a loss of offsite power. The inspector expressed concern to licensee management over the potential for developing Reactor Coolant Pump seal degradation during this situation. The licensee representative stated that an evaluation would be performed of the potential consequences of securing both CCW and seal injection during an event requiring safety injection (e.g. steam generator tube rupture) con-current with a loss of offsite power. A schedule for performing the evaluation and developing recommendations, if necessary, for hardware and/or procedure changes shall be developed by mid-July,1982.
6.
Material Accountability for Steam Generator Modifications Reference: NRC Region I Inspection Reports 82-06 and 82-07 a)
The inspector reviewed completed Quality Control (QC) Surveillance Reports dealing with material accountability during the modification work in both steam generators. The following Surveillance Reports were reviewed:
82-228-310-345-286-31 5-346-306-341-361 Eact report discussed objects for which retrieval or disposition was necessary.
The objects were either known to have been accidentially dropped from the work platform during the modification or were discovered during tubesheet video inspection by the licensee. The QC inspectors documented the dropped or dis-covered objects; however, they checked a block on the preprinted Surveillance Report form stating that corrective action was not required.
Further,QC supervision reviewed, signed and closed out the above Reports without document-ing what actions were taken with regard to removing the objects. Discussions
l with QC supervision indicated that, although not documented on the Surveillance Reports, in all but one case (82-310), they were aware the objects were removed or properly dispositioned prior to signing and closing out the respective Sur-veillance Reports. Surveillance Report 82-310 documented that a weld rod was i
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dropped inside the 'B' Steam Generator during modifications, with no corrective actions indicated as required. QC supervision reviewed and closed out the Report four days prior to removal of the weld rod.
Fail-ure to properly document the need for corrective action and then track Report 82-310 through completion of follow-up actions is contrary to 10 CFR 50, Appendix B, Criterion X and Ginna Administrative Procedure-1001, Inspection and Surveillance Activities, and is considered a violation (82-10-01).
b.
During several periods between May 12-20, 1982, the inspectors reviewed final video taped inspections from both the 'A'
and 'B' Steam Generators.
Included in the review were:
'A' Steam Generator final tubesheet inspection of columns 21 through
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38-hot and cold leg side (viewed from the blowdown lane looking down columns).
'A' Steam Generator tubesheet inspection of peripheral tubes R40C66
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thru R45C39-hot leg side (viewed from periphery at right angles).
(viewed from periphery looking forward)pection of entire periphery
' A' Steam Generator final tubesheet ins
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'B' Steam Generator final tubesheet inspection of columns 20 through 39-hot and cold leg side (viewed from the blowdown lane looking down columns).
'B' Steam Generator final tubesheet inspection of entire periphery
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(viewed from periphery looking forward).
'B' Steam Generator general inspection of the tube stabilization
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modification performed in the 'U' bend region.
Review of the video inspections did not identify any foreign objects of a mass which was considered to have the potential to be detrimental to the steam generators. Very small pieces of debris approximately inch diameter or less were noted to a limited extent in both steam generators. Several small pieces of weld rod were also identified in the 'A' Steam Generator and were not considered detrimental.
c.
During the outage the licensee installed a metal impact monitoring system designed to detect the presence of loose objects in the steam generators.
The system consists of four accelerometers mechanically mounted to the outside surface of the steam generators. Two accelerometers are located 900 apart on the tubesheet centerline elevation, and the remaining accelerometers are located 2h feet above and below the tubesheet centerline.
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During initial startup following the outage and with the reactor critical at powers less than 5%, several alarms were received on 'B' Steam Gener-ator from the newly installed system.
Evaluation by the licensee and the vendor, Westinghouse, determined the source of the impacts to be at an elevation in the vicinity of the fifth support plate; however, it was not determined whether the source was originating from inside or outside the steam generator.
It was noted that a snubber ring header is located at that same general elevation. Additional temporary accelerometers were attached to the main steam line, feedwater line, and lifting eyes on the 'B' Steam Generator to better pinpoint the origin of the impacts.
Power was increased to 25% with no further alarms received. The licensee reduced power to approx-imately 1% to try and recreate the alarm condition. Additional impacts at about 1% power were received, evaluated and determined to be originating on the cold leg side at about the same elevation as previously determined.
Following power increase to 25% with no further alarms being received, a normal physics testing startup to full power was resumed, and again, no alarms occurred.
Based on the absence of alarms except at very low power, a loose part was not considered to be the originating source of the impacts. A hydraulic snubber, located near the fifth support plate on the cold leg side is be-lieved to be one possible source of the impacts, due to thermal-mechanical interaction between the snubber support and its mounting on the shell of the steam generator during heatup/cooldown of the secondary side. Also under consideration is a thermally-induced condition inside the steam generator brought about through system heatup/cooldown, which at low power may be affect-ed by the much cooler auxiliary feedwater flow. The inspector will continue
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7.
Inspector Witnessing of Surveillance Test The inspector witnessed the performance of surveillance testing of selected a.
components to verify that the surveillance test procedure was properly ap-proved and in use; test instrumentation required by the procedure was cali-brated and in use; Technical Specifications were satisfied prior to removal of the system from service; test was performed by qualified personnel; the procedure was adequately detailed to assure performance of a satisfactory surveillance; and test results satisfied the procedural acceptance criteria, or were properly dispositioned.
b.
The inspector witnessed the performance of:
Refueling Shutdown Surveillance Procedure (RSSP)-6.0, Containment
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Integrated Leakage Rate, Revision 9, April 24,1982. Portions of the test were observed during May 3-5, 1982.
RSSP-15.12, Hydro Test of Class 'B' Piping (A and/or B S/G Secondary
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Piping), Revision 3, May 13, 1982, performed May 14, 1982.
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Periodic Test (PT)-7, Hydro Test of Reactor Coolant System, Revision
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24, December 23, 1980, performed May 20, 1982.
c.
The Containment Integrated Leakage Test was temporarily stopped shortly after pressurization due to excessive leakage from the containment purge supply and exhaust valves. The supply and exhaust valves outside contain-ment were adjusted and the test resumed without further complication.
Initial review of data indicates a measured leakage rate of approximately.0076 weight percent /24 hours.
In accordance with Technical Specification 4.4.16 the licensee will submit a sumary technical report to the NRC.
8.
Inspector Witnessing of Plant Maintenance and Modifications a.
During the inspection period, the inspector observed various maintenance and problem investigation activities to verify compliance with regulatory requirements, including those stated in the Technical Specifications; com-pliance with administrative and maintenance procedures; compliance with applicable codes and standards; required QA/QC involvement; proper use of safety tags; proper equipment alignment and use of jumpers; personnel quali-fications; radiological controls for worker protection; retest requirements; and ascertain reportability as required by Technical Specifications'.
In a similiar manner the implementation of design changes and modifications were reviewed. Compliance with requirements to update procedures and drawings were verified and post modification acceptance testing was evaluated.
b.
The inspector witnessed the following maintenance activity:
Repair of purge supply and exhaust valves, perfomed May 18, 1982.
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c.
Following completion of the Containment Integrated Leakage Test, with contain--
ment still pressurized to 35 psig, a small b" test line between the inner and outer containment purge supply valves was opened to determine local leak-age in the accident direction. Leakage of approximately 200,000 cc/ minute was measured past the inner valve. The routine, quarterly test, which pressurizes the volume between the valves to 60 psig was then conducted. This test, which pressurizes the opposite side of the inner valve, measured leakage of only 67 cc/ minute. Similiar, although not as dramatic differences in leakage were found on the purge exhaust inner isolation valve. The purge valves are 48" butterfly valves with resilient valve seats. With a Henry Pratt Company rep-resentative present, it was determined that the cause of the directional leak-age was due in part from the valve disc not coming to rest in the center of the seat upon closing. Additionally, the seats had not been replaced since original installation and proper adjustment between the seating surface and discs, as verified by feeler guage, had not been previously incorporated into the appropriate maintenance procedure. Maintenance Procedure (M)-37.15 was revised to include the proper method for assuring the correct interference between the disc edge and seat. The seats associated with the four purge valves were replaced and local leak rate testing was satisfactorily performed on May 17 and 19,198..
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Review of Periodic and Special Reports ~
Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specification 6.9.ii and 6.9.3 were reviewed by the inspector. This re-view included the following c.onsiderations:.the report included the information required to be reported by FtC requirements;mtest results and/or supporting in-formation were consistent with design predictions and performance specifications; planned corrective action wa 5 adequate for resolution of identified prcblbms; determination whether any ir. formation in the report required classification as an abnormal occurrence; and the validity of the' reported inform'ation. Within the. ~~ ' -
scope of the above, the following periodic report was reviewed by the inspector.'
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Monthly Operating Report for April,1982.'
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Exit Interview
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At periodic intervals durint the course of the inspection, meetings werc' held with senior facility manager;ent to discuss the inspection scope and. findings..
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