IR 05000335/1981035
| ML17212B512 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 02/12/1982 |
| From: | Bibb H, Dance H, Elrod S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17212B509 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-3.D.3.4, TASK-TM 50-335-81-35, 50-389-81-26, IEB-78-12A, IEB-78-12B, IEB-78-13, IEB-79-27, NUDOCS 8204140215 | |
| Download: ML17212B512 (14) | |
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UNITEDSTATES NUCLEAR REGULATORY COMMISSlON
REGION II
'101 MARIETTAST;, N.W., SUITE 3100 ATLANTA,GEORGIA 30303 Repor t No. 50-335/81-35 and 50-389/81-26 Licensee:
Florida Power and Light Company 9250 West Flagler Street Miami, FL 33152 Facility Name:
St.
Lucie Units No.l and No.
Docket No. 50-335 and 50-389 License No'.
DPR-67 CPPR-144 Inspection at St. Lucie site near Ft. Pierce, Florida Inspectors:
S.
A. Elrod Da e igned H.
E. Bibb Approved by:
H.
C.
Dance, Section C ref, Division of Resident and Reactor Project Inspection SUhlhNRY Inspection on December ll, 1981 - January 10, 1982
Da e Si ned t
Signed Areas Inspected This routine, inspection involved 164 resident"inspector-hours on site in the areas of Haintenance Observation, Technical Specification Surveillances, Licensee Event Reports Review, Bulletins, Informhtion Notices, Operational Safety Verification and Tel Action Plan Items.
Results Of the areas inspected, one apparent item of noncompliance was identified in
area; no apparent items of noncompliance were found in 6 areas.
82041402i5 820330 PDR ADDCK 00000335
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DETAILS Persons Contacted Licensee Empl oyees
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M. Wethy, Plant Manager H. Barrow, Operations Superintendent E.
Bowers, Maintenance Superintendent A. Sager, Operations Supervisor G.
Roos, guality Control Supervisor J. Frechette, Chemistry Supervisor F. Leppla, Instrument and Control Supervisor L. Fincher, Training Su perv isor R. Jennings, Technical Department Supervisor W. Mikell, Outage Coordinator A. Pell, Reactor Engineering Supervisor F. Buchanan, Health Physics Supervisor G. West, Security Supervisor Barrow, Fire Prevention Coordinator D. Hayes, Nuclear Plant Supervisor W. Pearce, Nuclear Plant Supervisor D. West, Nuclear Plant Supervisor L. Bur ton, Nuclear Plant Supervisor B. Vincent, Assistant Plant Superintendent-W. Bailey, guality Assurance Supervisor Electrical Other licensee employees contacted included construction craftsmen, technicians, operators, mechanics, security force members, and office personnel.
- Attended exit interview Exit Interview The inspection scope and findings were summarized on January 2, 1982, with those persons indicated in paragraph 1 above.
Licensee Action on Previous Inspection Findings Not inspected.
Unresol ved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve noncompliance or deviations.
New unresolved items identified during this inspection are discussed in paragraph 1 ll I II
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Technical Specification Surveillances During the inspection period, the inspector conducted a comprehensive audit of procedural compliance with technical specification surveillance requirements.
Subsequently, 62 surveillance requirements have been researched.
These include paragraphs 4.1.1.1.1.
a through 4.3.3.7.1 inclusive.
For each requirement, a written procedure was reviewed and found to be current and accurate with the technical content sufficient to meet the intent of the surveillance requirement.
No violations or deviations were identified in this area.
IE Bulletins The following IE Bulletins were reviewed to determine whether they had been received and reviewed by appropriate management, responses, where necessary, were accurate and complete, and that action taken was complete.
(Closed - Unit 1)
IEB 79-27 Loss of None-Class IE Instrumentation and Control Power System Bus During Operation The inspector reviewed the licensee's response to IEB 79-27 and was satisfied that the bulletin requir ements had been met.
Additionally, the following procedures were reviewed to assure that loss of power situations had been adequately addressed:
EOP 0030140 EOP 0030142 EOP 0040030 EOP 0910030 EOP 0910031 EOP 0910032 (Closed-Unit 2)
Blackout Operation RCS Cooldown During Blackout SDC/LPSI Off-Normal Operation Start-up Transformer Off-Normal Operation tlain Transformer Off-Normal Operation Auxiliary Transformer Off-Normal Operation IEB 78-13, 12A, 12B - Atypical Weld Material in Reactor Pressure Vessel Welds.
These IE Bulletins are closed based on IE:Hg review of submittals from the manufacturer, Combustion Engineering Company.
IE Information Notices The following IE Information Notices were reviewed to ensure their receipt and review by appropriate management.
IE Information notices are considered closed upon receipt and review.
IEN 81-38:
Potentially Significant Equipment Failures Resulting From Contamination of Air-operated Systems
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The following IE Information Notice was sent only to byproduct material licensees and is closed as not applicable to St. Lucie.
IEN 81-37 Unnecessary Radiation Exposure to Public and Workers during Events Involving Thickness and Level Measuring Devices 8.
Licensee Event Reports Review The following LER's were reviewed to verify that reporting requirements had been met, causes had been identified, corrective action appears appropriate, generic applicability had been considered, and the LER forms were complete.
Additionally, for those reports identified by asterisk, a more detailed review was performed to verify that the licensee had reviewed the events, corrective action had been taken, no unreviewed safety questions were involved, and violations of regulations or Technical Specification conditions had been identified.
- LER 81-47
"A" Diesel Generator Turbocharger
- LER 81-48 Seismic Instrumentation 9.
Licensee Actions taken with Respect to the Th11 Action Plan (NUREG-0737)
a.
(Closed-Unit 1)
TAP III. D.3.4 Control Room Habitability Licensee analysis and submittal L-81-4 dated January 2, 1981 concludes that no further action is necessary because previous review and modifications had upgraded the system to comply with Standard Review Plan 2.2. 1, 2.2.2, 2.2.3 and 6.4.
Within L-81-4, sections of the St.
Lucie
FSAR are included because the hazards analysis for St. Lucie 2 also applies to Unit 1.
Both control rooms have been evaluated recently - Unit 2 in NUREG 0843, the Safety Evaluation Report for St.
Lucie 2 operating license, and Unit 1 in inspection report 335/81-13, paragraph 4.1. 1. 1.
Based on these reviews, the inspector concludes that TAP III.D.3.4 is satisfied.
This item is closed.
b.
(Open) Unit 1 TAP II.B.1 Reactor Coolant System Vents This system was installed during the fall 1981 refueling outage in accordance with plant change/modification 56-80 as described to NRC in FP8L letter L-81-347 dated August 10, 1981.
When tested at pressure in accordance with preoperational,test procedure 0120086, Rev. 0, all three downstream valves opened when any of the four upstream valves were actuated.
This test was witnessed by the inspector.
That these valves opened appears to be related to the use of solenoid pilot valves.
The system was manually isolated at the reactor and at the pressurizer and was electrically de-energized while design studies continue.
It was noted by the inspector that the procedures for use of this system have not been developed.
This was the subject of correspondence
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dated November 30, 1981 between NRC (NRR) and the licensee.
This item remains open pending completion of design studies and modification of the present installation.
Maintenance Observation Station Maintenance activities of safety-related systems and components listed below were observed/reviewed to ascertain that they were conducted in accordance with requirements.
The following items were considered during this review:
the limiting conditions for operation were met, activities were accomplised using approved procedures, functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintianed; activities were accomplished by 'qualified personnel; parts and materials used were properly certified; and radiological controls were implemented as required.
Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may affect system performance.
The following maintenance activities were observed/reviewed:
On December 19, 1981, during recovery from a reactor trip, paragraph llc, a pressurizer code safety relief valve V-1200, lifted at a pressure of 2250 psig instead of the required setpoint of 2500'5.
At the time of lift, workers were completing replacement of a rupture disc on the quench tank, the tank to which the pressurizer power operated relief valves and code safety relief valves discharge upon lifting.
The quench tank rupture disc had been ruptured some hours earlier during a complete loss of load transient.
The workers heard the valve begin relieving to the tank and hurriedly exited the immediate area before the disc ruptured for the second time.
No one was injured, and about 1500 gallons of primary water was lost to the sump.
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The valve, V-1200, was removed and replaced with a spare.
Some days later (12/29/81) the valve was tested and the nozzle ring adjustment setpoint was found in the "as is" condition to be one notch from the disc contact point.
Two test lifts at this setting showed the valve to liftat 2250 and 2200 psig.
Turning the nozzle ring to 8 notches from the disc contact point (as recommended by manufacturer for bench testing)
caused a
lift point of 2350 psig on two tests.
This testing was performed under plant work order 2637.
Plant Work Order 3538 (completed 4/80)
was reviewed to obtain background information on pressurizer code safety relief valve, V-1200.
Paragraph 9. 1.8 of the Maintenance Procedure M-0017 Rev.
0 showed the nozzle ring setting to be 14 from the disc contact point notches prior to testing the valve.
This is the correct nozzle ring setting's stamped on the bonnet flange for this valve.
Paragraph 3.2.4 of Crosby Valve Company Technical Manual I-1105-2 indicates this to be the correct procedure.
However, during bench tqsting of the valve, the nozzle ring gets set at 8 notches from the disc ring contact per paragraph 9.1.9 of the procedure M-0017.
Upon completion of testing, paragraph 9. 1. 19 of the procedure
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indicates that the nozzle ring setting should be returned to the original setting found in paragraph 9.1.8.
It would appear that the mechanic rotated the nozzle ring in the wrong direction from the i"~ test setting and ended up at 1 notch from the disc contact point. in lieu of 14 notches.
C The maintenance procedure, t10017, contains several g.C.
"Hold" points for independent verification signoff, but paragraphs which involve nozzle ring settings are not among these.
Additionally, upon disassembly of the valve it was noted that both the adjusting ring and the nozzle ring set screws had been improperly installed after the previous testing.
Deep indentations were present on a tooth land on each ring and the set screw point was damaged.
Paragraph 8.2.11 cautions one to be sure the set screw point engages a notch.
Tightening down the set screw on a land puts a strong side force on the adjusting and nozzle rings distorting the seat position after the valve is lifted one time.
Licensee conversations with the valve vendor have indicated that this was the most probable cause of premature lift the second time.
Failure to include the requirement to install the set screw in a notch and failure to follow the existing procedure by turning the nozzle rings the wrong way are examples 'of a violation of technical specification 6.8. 1 (335/81-35-01).
During this inspection, it was discovered that parts have been exchanged between code safety valves.
The vendor technical manual at paragraph 3.2.3.3 cautions that the bonnet assembly, nozzle ring and set screw are identified and matched with a specific valve body by tag number and should be assembled accordingly.
The licensee indicates that the vendor has stated via telephone that the reason for that caution is not clear.
The accept-ability of swapping the above relief valve parts is unresolved pending formal clarification of the technical manual (335/81-35-02).
11.
Operational Safety Verification The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the report period.
The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components.
Tours of the reactor auxiliary and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance.
The inspector by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security plan.
b.
The inspector observed plant housekeeping/cleanliness conditions and verified implementation of radiation protection controls.
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inspector also witnessed portions of the radioactive waste system controls associated with radwaste shipments.
Reactor Trip from Hain Steam Isolation Valve (HSIV) Closure On December 19, 1981 Unit 1 tripped at 4:14 a.m.
due to main steam isolation valve closure causing a complete loss of load.
At the time the pressurizer power operated relief valves opened to prevent over pressure.
They were subsequently isolated because pressure dropped to 1650 psig.
The reactor coolant system was repressurized and the valves unisolated.
Since the valves were seated, reactor startup was commenced.
Later, while completing re-installation of the pressurizer relief tank rupture disc, the code safety, V-1200, lifted at 2250 psig and blew down to 1650 psig.
The reactor tripped, and the relief tank rupture disc ruptured again.
The reactor was placed in cold shutdown to repair the code safety.
The code safety repair is further discussed in paragraph 10.
The licensee and Architect/Engineer investigated the closure of the HSIV's extensively.
No positive cause was identified, however the air system that holds the valves open was modified by plant change/modi-fication 002-82 to add two redundant air compressors and backup air cylinders operating at approximately 105 psig.
This modification is intended to increase the reliability of the air supply and to increase the operating air pressure to ensure the valves are held open under the higher steam flows associated with a recently approved power level increas Ue Uf Ut l
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