IR 05000244/1986003
| ML17251A624 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/25/1986 |
| From: | Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17251A621 | List: |
| References | |
| 50-244-86-03, 50-244-86-3, NUDOCS 8604040364 | |
| Download: ML17251A624 (16) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-244/86-03 Oocket No. 50-244 Licensee No.
OPR-18 Priority Category C
Licensee:
Rochester Gas and Electric Corporation 49 East Avenue Rochester, New York 14649 Facility Name:
R.
E. Ginna Nuclear Power Plant Inspection at:
Ontario, New York Inspection Conducted:
February 9, 1986 through February 28, 1986 Inspectors:
W. A. Cook, Senior Resident Inspector, Ginna T.
K. Kim, Resident Inspector (Trainee),
Ginna Approved by:
J Linville, C ie, Reactor P oject Sectio
.
2C, ORP at Ins ection Summar
Ins ection on Februar
1986 throu h Februar
1986 Re ort No.
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Areas Ins ected:
Routine, onsite, regular, and backshift inspection by the resident inspectors ( 113 hours0.00131 days <br />0.0314 hours <br />1.868386e-4 weeks <br />4.29965e-5 months <br />).
Areas inspected included: plant operations; licensee action on previous findings; surveillance testing; maintenance; refueling activities; Licensee Event Reports; review of periodic and special reports; and inspection of accessible portions of the facility during plant tours.
Results:
In the seven areas inspected, one violation was identified.
A violation of refueling containment integrity requirements is discussed in paragraph 3a.
A review of refueling operations and guide tube split pin replacement is discussed in paragraph 6.
8604040364 860328 PDR ADOCK 05000244'
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DETAILS 1.
Persons Contacted During this inspection period, the inspector interviewed and talked with operators, technicians, engineering and supervisory level personnel.
2.
Licensee Action on Previous Ins ection Findin s
During an NRC team inspection conducted January 27-31, 1986, (Inspection Report 50-244/86-02)
the inspectors determined that the licensee was not performing Nondestructive Examination (NDE) on the reactor vessel head lifting rig and internals lifting-rig as committed to in a RGKE to NRC letter, dated March 2, 1983.
Further review determined that licensee personnel required to perform the NDE and those responsible for verification of these activities were not fully cognizant of the requirements.
The licensee took immediate action t'o correct these deficiencies prior to lifting the reactor vessel head and internals package during the upcoming 1986 Refueling Outage.
The inspectors verified that Materials Handling Equipment Procedure No.
MHE-1000-1, "Inspection and Maintenance of Special Lifting Devices",
Quality Maintenance Procedure No.
QM 1315, "Inspection of Lifting Devices in the Containment Vessel",
and Ginna Station Refueling Procedure (RF)-61, "Cycle XV-XVI Refueling Procedure",
were properly revised and that responsible licensee personnel were trained and cognizant of the lifting rig NDE requirements.
The inspectors witnessed a portion of the NDE of the lifting rigs and reviewed the examination results upon completion of the inspections on February 10, 1986.
No discrepancies were noted.
In addition, the inspectors observed a
presentation given by senior licensee management on February 6, 1986 to station personnel emphasizing the importance'of procedural compliance, quality assurance and quality control in all station activities.
3.
Review of Plant 0 erations a
~
Throughout the reporting period, the inspectors reviewed refueling outage activities.
The reactor was shutdown on February 8, 1986 and has been in cold shutdown the entire inspection period.
Major activities observed by the inspectors include:
control rod guide tube split pin replacement; refueling; reactor disassembly and re-assembly; steam generator nozzle dam installation; Eddy Current testing; tube sleeving and plugging; numerous motor and valve repairs; and modification work.
On February 15, 1986, refueling operations were being conducted in accordance with Refueling Procedure (RF)-61, "Cycle XV-XVI Refueling Procedure".
At approximately 3:30 P.M., the licensee deenergized Motor Control Center (MCC)
1C and Vital Bus 14 to perform scheduled preventive maintenance.
This maintenance activity continued into February 16, 198 V I
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During the first Operations shift (0-8) of February 16, control room operators performed Refueling Procedure (RF)-8.2,
"Fuel Handling Instruction Pre-loading and Periodic Valve Alignment Check", to verify Technical Specification refueling requirements were satisfied.
At 1: 15 A.M., operators identified that motor-operated valve (MOV) 313 and MOV 813 (containment isolation valves for reactor coolant pump seal water return and reactor vessel support component cooling water return, respectively)
were inoperable because power was removed from their operators.
MOV 313 and MOV 813 are supplied power by MCC 1C which is fed power from Vital Bus 14.
Automatic containment isolation MOVs 313 and 813 are required by Technical Specifications to be operable or at least have one valve in the same line locked closed to meet containment integrity requirements for refueling operations.
Upon identification that the valves were inoperable, the licensee ceased refueling activities.
Both MOV 313 and MOV 813 were determined to be deenergized in the open position.
To satisfy Technical Specification 3.8 '.a.,
Operations personnel closed and'agged closed two manual isolation valves (V-315A and V-315C)
downstream of MOV 313 and manually closed and tagged MOV 813.
Neither containment isolation valve provides a direct communication path from inside containment to the outside atmosphere.
Both valves are associated with closed systems which penetrate containment.
The licensee resumed refueling operations upon completion of compensatory measures and reverification of refueling Technical Specifications.
The inspector verified that the licensee made the appropriate notification to the NRC Headquarters duty officer via the Emergency Notification System.
The inspector determined that this was the first outage that the licensee performed maintenance on Vital Bus 14 and MCC 1C concurrent with refueling operations.
The maintenance procedures used to establish the necessary isolation to perform this work do not caution against the potential impact on refueling operations.
The inspector determined that the review conducted for scheduling this maintenance activity did not include a comprehensive review of all electrical loads supplied by MCC 1C.
However, the inoperability of MOV 313 and MOV 813 was identified by onshift licensed operators during routine verification of containment integrity, performed once a day while core alterations are being conducted.
Maintenance performed on Vital Bus 14 and MCC 1C resulted in a violation of refueling containment integrity requirements on February 15 and 16, 1986.
Although the inoperable valves were identified and reported by the licensee, and of relatively minor safety significance; Technical Specification 6.8, Regulatory Guide 1.33, Appendix A, and ANSI N18.7-10172 state that planning of main-tenance shall consider the possible safety consequences of
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concurrent or sequential maintenance, testing or operating activities.
In that this is the second recent example of station maintenance activities impacting upon the operability of Technical Specification required equipment, this activity is a violation.
(86-03-01)
b.
During the inspection, accessible plant areas were toured.
Items reviewed include radiation protection and contamination controls, plant housekeeping, fire protection, equipment tagging, personnel safety, and security.
On February 11, 1986, the licensee's control room simulator arrived onsite from Westinghouse after completion of factory acceptance testing in Pittsburgh.
The licensee is completing the final phases of their acceptance testing program and plans to conduct plant start-up training classes the week of Harch 10, 1986.
The licensee plans to conduct additional operator training on plant start-up transients after having experienced multiple operator-related plant control problems during start-up from the 1985 Refueling Outage.
In addition, the inspector determined that the licensee currently plans to start symptom-oriented Emergency Operating Procedures training on the new simulator in April 1986.
-Inspector tours of the control room this inspection period included reviews of shift manning, operating logs and records, equipment 'and monitoring instrumentation status.
d.
Safety system valves and electrical breakers were verified to be in the position or condition required for the applicable plant mode as specified by Technical Specifications and plant lineup procedures.
This verification included routine control board indication review and conduct of a partial systems lineup check of the Residual Heat Removal System on February 25 and 26, and the 1B Emergency Diesel Generator on February 27, 1986.
4.
Survei 1 lance Testin a.
The inspector witnessed the performance of surveillance testing of selected components to verify that the test procedure was properly approved and adequately detailed to assure performance of a satis-factory surveillance test; test instrumentation required by the procedure was calibrated and in use; the test was performed by qualified personnel; and the test results satisfied Technical Specifications and procedural acceptance criteria, or were properly resolved.
b.
The inspector witnessed the performance of a portion of the following tests:
Periodic Test (PT)-2.3. 1,,"Post Accident Charcoal Filter Dampers",
Revision 12, dated January 28, 1986, performed February 11, 1986.
Refueling Shutdown Surveillance Procedure (RSSP)-ll, "Pressurizer Safety Valve Test", Revision 10, dated May 25, 1984 performed February 20, 1986 on relief valve RV-434.
RSSP-15. 1, "Hydrostatic test on Class B Safety-Related Piping (Safety Injection & Containment Spray Suction Piping, to include NaOH piping)", Revision 6, dated August 23, 1984, preparations reviewed February 21, 1986.
SM-4136.5, "Run-testing of the A Diesel Generator from the Emer-gency Local Control Panel",
Revision 0, dated February 16, 1986, testing of local control features at the newly installed Local Emergency Control-Panel on February 27, 1986.
RSSP-2.3,
"Emergency Diesel Generator Trip Testing", Revision 20, dated February 26, 1986, overspeed trip testing performed February 27,1986.
No violations were observed.
5.
Plant Maintenance a.
Dur in the i g
nspection period, the inspector observed maintenance and problem investigation activities to verify compliance with regulatory requirements, including those stated in the Technical Specifications; compliance with administrative and maintenance procedures; required gA/gC involvement; proper use of safety tags; proper equipment alignment and use of jumpers; personnel qualifications; radiological controls for workers protection; and reportability as required by Technical Specifications.
b.
The inspector witnessed portions of the following maintenance activities:
Preparations for removal of RCP seal water injection filters in accordance with Maintenance Procedure (M)-7.2, "Seal Water Injection Filter Replacement",
Revision 7, dated June 18, 1984, observed on February 12, 1986.
Preparations for installation of nozzle dams in the A steam generator cold leg in accordance with M-43.44. 1, "Installation of Steam Generator Nozzle Dam, A Steam Generator",
Revision 6, dated October 8, 1985, observed on February 12, 1986.
Installation of pressurizer safety valves in accordance with M-37.2.2, "Inspection and Maintenance of Pressurizer Safety Valves
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RV-434 or RV-435", Revision 7, dated August 10, 1984, observed on February 27, 1986.
Maintenance performed on the 1A component cooling water pump in accordance with M-11. 14, "Inspection and Maintenance of Ingersol Rand Pumps",
Revision 13, dated December 20, 1985, observed February 21, 1986.
Calibration of A steam generator wide range level transmitter, L-460, in accordance with Calibration Procedure (CP)-460,
"Calibration and/or Maintenance of Steam Generator A Wide Range Level Channel L-460", Revision 4, dated December 10, 1985, observed on February 27, 1986.
Calibration of volume control tank level transmitter channel 112 in accordance with CP-112, "Calibration and/or Maintenance of Volume Control Tank Level Channel 112", Revision 1, dated May 8, 1984, observed on February 21, 1986.
Calibration of RCS accumulator level channel 935 in accordance with CP-935, "Calibration and/or Maintenance of Accumulator Level Channel 935", Revision 0, dated February 8, 1983, observed on February 20, 1986.
Cable splicing of electrical leads for Appendix R modification inside containment in accordance with Station Modification Procedure (SM)-4134.3,
"Appendix R Instrumentation Installation Inside Containment",
Revision 0, dated February 6,
1986, observed on February 28, 1986.
Installation of purge system flanges and davit assemblies inside containment in accordance with SM-2504.3,
"Removal of 48" Purge System Supply Valve and Blind Flange Installation Inside Containment at Penetration 204", Revision 1, dated February 19, 1986, observed periodically between February 18 and 27, 1986.
No violations were observed.
6.
Refuelin Activities The inspectors observed refueling activities conducted in accordance with Refueling Procedure (RF)-61, "Cycle XV-XVI Refueling Procedure",
Revision 0, dated February 8, 1986.
The inspector verified that RF-61 was properly reviewed and approved, technically adequate and in compliance with Station Technical Specifications'n addition, the inspectors reviewed refueling checklists RF-8.2,
"Fuel Handling Instructions Pre-loading and Periodic Valve Alignment Check", RF-8.3,
"Core Loading Periodic Equipment Status Check",
and pre-refueling checklists contained in RF-61 to verify that necessary plant conditions,
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systems, radiation monitors, instrumentation, handling equipment, communications and personnel requirements stipulated by station procedures and Technical Specification were satisfied prior to and during refueling operations.
The inspectors witnessed fuel movement activities on February 15 and February 26, 1986 from the refueling floor in containment, the control room and the auxiliary building spent fuel pit area.
Refueling operations were performed by a team of Westinghouse contractors.
During refueling operations the inspectors observed that all core alterations were directly supervised by a licensee Senior Reactor Operator.
Communication was maintained via headphones between the refueling team in containment, the control room and the spent fuel pit.
All fuel movements were recorded on the refuel floor and the control room.
Small tools and equipment on the refuel floor were properly controlled or tethered.
In addition, the inspectors observed that licensee guaality Control inspectors provided continuous coverage of refueling activities.
Based on a review of completed records, discussion with control room operators and personal observations, the inspectors verified Technical Specifications for fuel handling were properly satisfied.
This included verification that:
the minimum number of source range monitors were operable and one source range monitor provided audible indication in containment; refueling cavity water level was maintained greater than
feet above the reactor vessel flange; primary coolant system boron concentration was maintained greater than 2000 ppm; a minimum of one Residual Heat Removal loop was in operation; containment radiation levels were continuously monitored; containment and auxiliary building ventilation systems were properly aligned; and containment integrity requirements were maintained.
One event resulting in a containment integrity violation during refueling operations is discussed in paragraph 3.a.
Inspectors monitored the split pin replacement work performed by Framatome.
Framatome was contracted by the licensee to conduct control rod drive guide tube split pin replacements during this refueling outage.
The split pins were replaced due to their susceptibility to failure and potential entrainment of loose parts in reactor coolant flow potentially resulting in mechanical fouling.
(see IE Information Notice No. 82-29)
Of the 33 guide tubes refurbished, Framatome discovered one failed split pin in guide tube G-3.
The fai lure is believed to have been caused by stress corrosion cracking in the shank area of the pin right below the lock nut engagement.
The lower segment of the failed pin was safely retrieved from the upper core plate of the internals package.
All of the. new split pins are solution heat treated at higher temperatures than the old pins and the torque on the lock nut is reduced to prevent premature stress corrosion crackin After split pin.replacements, Framatome's guality Control group conducted video inspections of all the guide tubes and the upper internals package to insure control rod guide tubes were not damaged and aligned properly.
In addition to the split pin replacement, the licensee, through a Framatome subcontractor, replaced the existing guide tube flexure inserts with new flexureless inserts.
The inserts seat in the top of the guide tubes and provide support to the control rod drive shaft as it moves up and down through the guide tube.
The inspectors observed a portion of the core mapping sequence on February 27, 1986 and discussed the core verification procedure with licensee and Westinghouse representatives.
The core map was videotaped by the Westinghouse team.
A positive identification was made of each fuel assembly by a Westinghouse representative and simultaneously verified by a licensee gC inspector, the refueling SRO and another Westinghouse team member in the control room.. The videotapes were later reviewed separately by the licensee's Reactor Engineer.
No discrepan-cies were noted.
Licensee Event Re orts LERs The inspector reviewed the following LER to verify that the details of the event were clearly reported, the description of the cause was accurate, and adequate corrective action was taken.
The inspector also determined whether further information was required, and whether generic implications were involved.
The inspector further verified that the reporting requirements of Technical Specifications and station administrative and operating procedures had been met; that the event was reviewed by the Plant Operations Review Committee and that continued operation of the facility was conducted within the Technical Specification limits.
86-.01: "Failure to Meet Minimum Oegree of Redundancy for Engineered Safety Features Actuation System".
On January 18, 1986 with reactor power at approximately 93 percent, a station Results and Test technician conducted inservice valve testing (stroking close and then open)
on each of the containment pressure transmitter manual isolation valves, one at a time.
The valve testing was performed in accordance with Periodic Test Procedure (PT)-2. 10.6,
"Containment Pressure Transmitters Manual Isolation Valve Stroking".
Upon subsequent review of the procedure by the licensee, it was determined that the isolation valves were stroked contrary to the requirements of the (juality Assurance Manual, Appendix C
and Technical Specifications.
Appendix C states that containment pressure transmitter isolation valves should be stroked only during cold shutdown or refueling.
PT-2. 10.6 did not, specify these initial conditions.
In addition, PT-2. 10..6 did not require that the containment pressure channel be defeated prior to stroking the isolation valves.
By not defeating the individual pressure channel, the minimum degree of redundancy requirement of Technical Specification Section 3.5.3.2 and
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Table 3.5-2 was not satisfied.
As a result one of the actuation signals to the Safety Injection, Containment Spray and Main Steam Isolation systems was momentarily compromised during the approximate ten seconds the transmitter isolation valve was being stroked.
A Notice of Violation is not issued for this event because the error was identified by the licensee and is of minor safety significance.
In addition, the licensee took prompt corrective action and reviewed similar guality Assurance Manual, Appendix C inservice valve testing procedures to ensure proper initial conditions are specified.
There have been no previous violations in this area.
8.
Review of Periodic and S ecial Re orts Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specification 6.9. 1 and 6.9.3 were reviewed by the inspector.
This review included the following considerations:
the reports contained the information required to be reported by NRC requirements; test results and/or supporting information were consistent with design predictions and performance specifications; and the validity of the reported information.
Within this scope, the following report was reviewed by the inspector:
Monthly Operating Report for January 1985.
9.
Exit Interview At periodic intervals and at the conclusion of the inspection period, meetings were held with senior facility management to discuss the inspection scope and findings.
Based on the NRC Region I review of this report and discussion held with licensee representatives, it was determined that this report does not contain information subject to
CFR 2.790 restriction /