ML20136D958
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i April 11, 1996-MEMORANDUM T0: Ellis W. Merschoff, Director l Division of Reactor Projects
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Albert F. Gibson, Director Division of Reactor Safety ;
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1 E. Imbro, Acting Director )
Project Directorate 11-2 , i Division of Reactor Projects I/II. I 1
FROM: Kerry D..Landis, Branch Chief- ,
Division of Reactor Projects i 4 ;
SUBJECT:
St. Lucie Unit 1 Refueling Outage ,
On April 29, St. Lucie Unit 1 is scheduled to shut down for a 47 day refueling ;
outage. The unit is scheduled to restart on June 9,1996. Enclosed is a matrix of major work activities and significant modifications to be accomplished during i the outage. Also enclosed is a schedule of critical path tasks.
This information is' provided to assist your staff 'in scheduling any. required j inspections or other regulatory actions during this. outage. For additional information or clarification you may contact Edwin Lea'at 331-3641.
Attachments:
- 1. Refueling Outage Items
.2. Critical Path Tasks i
i 1
I ncru-r RH HAP RH HAP A lf'. I SIGN AT URE / f NAME Eles:dka KLa ed d@lkg ;
DATF 04 / t / 96 ~041\ 196 04/l(/96 04 / 196 04 / /96 04 / /96 COPY 7 YES NO /fES') NO h NO YES NO YES NO YES NO OFFICIAL RECORD COPY W NT WAME P n0UTAGE.U1'
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31 g 7 970306 ] , l BINDEji96-485 PDR , 4 l
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. ATTACHMENT 1
- . l St. Lucie Unit 1 l Refueling Outage Items ACTIVITY REF. NRC PLANNED PLANNED LItENSEE LEAD / MODULE START COMP. CONTACT Verify Unit 1 PM SFP DRP Before Before C. Marple procedure for CLB RI/71707 4/29/96 4/29/96 x7294 ,
maintaining SFP review I temp <150F Verify SFP heat PM SFP DRP Before 'Before D. Denver i load calc CLB RI/37551 4/29/96 4/29/96 x7469 performed review Verify procedure PM SFP DRP Before Before C. Marple" to ensure boron CLB RI/71707 4/29/96 4/29/96 x7294 concentration review
>l720 ppm Unit shutdown DRP 4/28/96 4/29/69 C. Marple RI/71707 x7294 l RCS nozzle DRP 4/29/96 4/29/96 R. Riha inspections RI/71707 x7246
- Integrated DRP 5/2/96 5/3/96 R. Riha Safeguards RI/61726 x7246 l Testing Pzr code safety DRP 5/3/96 5/9/96 C. Ward valve RI/37551 x7275 replacement /flang
. e mods EDG Radiator DRP 5/6/96' 5/16/96 L. I l
Replacement RI/62703 5/17/96- 5/27/96 McLaughlin x3952 Visual Relief DRP 5/6/96 5/22/96 R.
examination of Request RI/62703 Frechette bottom of RWT. x3213 i
SG tube GL 95-03 DRS 5/6/96 5/22/96 R. Ball inspection. Bul. 89- x7109 100% bobbin 01 100%HL MRPC 3% CL MRPC -
100% Westinghouse Plug Rx Disassembly DRP 5/7/96' 5/16/96 R. Ball RI/62703 x7109 I
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.: , ATTACHMENT !
I St. Lucie Unit 1
] Refueling Outage Items
. Modify relief EEI 95- DRP 5/14/96 5/20/96 N. Motley
- i- valve setpoints 222 RI/92903 5/8/96 5/13/96 x7465 ,
2 and blowdown values - SDC
- j valves V3468 and V3483 l
Rx Vessel / Core DRS 5/16/96 5/21/96 K. Mayhew i Barrel inspection x7046 j
10 yr ISI
. Unit DRP 6/9/96 6/13/96 C. Marple',
- startup/ physics RI/71707 x7294 testing ]
4 Instrument LER DRP TBA TBA TBA sensing line 389/96- RI/62703
, blowdowns 01 I
EDG Relay socket LER DRP TBA TBA L.
replacements 389/95- RI/92902 McLaughlin j
05 x3952 o
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i' es-es-1we 07:e7Pr1 St Lucie Residsnt Offi i 407 461 4622 P.02 1
3 . i MID100P/ REDUCED IWWTORY CHECKLIST y
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Prior to plant operation with reactor coolant system partially drained (mid-
- loop), complete the following items
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l d I. Review Generic Letter 88-17.
l M. Ensure t e licensee has reviewed their controls and j administrative procedures governing mid-loop operation.
_ S. Procedures are active and in'use for the following '
requirements:
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- a. CONTAllMENT CLOSURE CAPABILITY FOR MITIGATION OF l RA0104CTIVE RELEASES '
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! # b. RC) TEMPERATURE'- AT LEAST TWO I E EPENDENT, CONTINU0US l IN91 CATIONS THAT ARE REPRESENTATIVE OF CORE EXIT
.i C0iDITIONS, ARE OPERABLE M. RCS LEVEL INDICATION - AT LEAST TWO INDEPENDENT.
j C0f'TINUOUS WATER LEVEL INDICATIONS OPERA 8LE (CALIBRATED)
! g. RCSPERTURBATIONSSHOULDBE5V0!DED i .
! gges. RCS INVENTORY ADDITION - AT LEAST TWO ADDITIONAL NEANS j
OF ADDING INVENT 0RY TO THE RCS MUST BE AVAILABLE. IN i ADDITION TO THE PUNPS THAT ARE PART OF THE NORMAL RHR
- ( SYETEMS - VERIFY OPERA 8ILITY ,
j M Y. N0HLE DANS/ LOOP STOP VALVES - REASONABLE ASSURANCE IS i OBTAINED THAT ALL NOT LESS ARE NOT BLOCKED i
SINULTANE0USLY UNLESS A VENT PATH IS'ESTA8LISHED TO PREVENT PRESSURIZATION OF THE UPPER PLENUM OF THE
- REACTOR VESSEL i
M. LICENSEE HAS CONTINGENCY' PLANS TO REP 0WER VITAL BUSSES FROM ALTERNATE SOURCE IF PRIMARY SOURCE IS LOST
- i 4. Fax a copy of this signed sheet to the appropriate Branch
- Chief upwi completion, j .
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Comments / Concerns: Ou r w e 9 b h -..a m AunNwrvr rn
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! Site: G L uc4Ef Inspect'or: _ E L Date: % '
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ATTACHNENT 1 4
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INSPECTOR INPUT St. Lucie i INPUT FOR INSPECTION REPORT NO.: 50-335,389/96-06 INSPECTOR: Rich C. Chou Reactor Inspector, Division of Reactor Safety DATES OF INSPECTION- May 6 - 10, 1996 i
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A. Inspector: l3 % % 5 M Ri h u, R tor Inspector Dite Si ed Approved : h-Paul Fredrickson, Chief
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. Ifato Signed Special Inspection Branch Division of Reactor Safety B. 1.0 Persons Contacted:
Licensee Employees '
B. Beisler, Structural Engineer ,
J. Brady, Shift Director i A. Fata, Civil Engineer !
C. Input for the appropriate inspection area.
3.0 Maintenance 3.1 Observation of Reactor Head Removal (62700)
The inspectors reviewed the adequacy of load testing for the p.olar crane to verify that the maximum lifting load would not exceed the maximum rated load and subsequently observed the reactor head lift. Procedures
, and documents reviewed for both the polar crane testing and reactor head lift were the following:
Condition Report 96-613, Evaluati.on of Polar Crane Load Test, Revision 0 Procedure 1-L01-MM-45, Unit 1 Reactor Containment Building Polar Crane Load Test of Main Hoist Gear Box to 125 percent of' Rated Load, Revision 4 Pro'cedure 1-M-0015, Reactor Vessel Maintenance - Sequence of Operations, Revision 27- '
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} The polar crane in the reactor building is used for the removal and }
- - installation of the reactor head and related parts. The original' rated' !
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- capacity for the polar crane was 350,000 lbs for the main hoist. :l Recently, the licensee identified problems with the polar crane gear box and decided to replace it. ANSI-B30.2 requires that the polar crane _ :
with the replaced gear box be retested for load capacity. Two tests ;
- . were performed.- In the first test, the vendor scale read-435,000 lbs - i l but the scale on the main hoist scale read 356,000 lbs. Because of the i discrepancy in scale readings, the licensee had the vendor ~ scale sent to J 3- a lab for verification. The vendor scale was determined to be reading !
too high and needed to be calibrated. The final calibrated load was 1 similar to the polar crane main hoist reading. In accordance with ANSI ;
~ B30.2, the' licensee used 80 percent (280,000 lbs) of the 350,000 lbs !
reading as the rated load. :
I LHowever, the licensee wanted to raise the rated loads for the crane and i a performed a second test. The second test was terminated prematurely i because a refilling hose for the water bag test load wa's broken. The i
! final reading for the scale before the hose broke was 357,180 lbs. The i
! licensee used 80 percent (285,744 lbs) of the tested load as the maximum -
- j. rated load to llft the reactor head. ,
In preparation of the reactor head lift, the licens'ee engineers reviewed
- the previous loads lifted for the reactor head and identified that the-
- loads ~were'between-280,000 and 320,000 lbs for the reactor head and I related items such as bolts, nuts, lead shielding, shielding support a frame, etc. The engineers also reviewed CE Instruction Manual No. 8770-i 12276, Rev. O entitled " Major Component Lifting / Lowering Interfaces for i
Reactor Vessel Head Upper Guide Structure, ICI plate, Core Support i-
' Barrel for the Florida' Power and Light Company, St. Lucie Unit 1 " which ,
stated that'the estimated load on the crane for the reactor vessel head
- lift was 267,968 lbs. From the sketches included in the manual, the licensee' engineers concluded that this weight ~ included the head and'the lift device (tripod) but did not include the studs or lead shielding. i
' The licensee engineers removed as many miscellaneous items as practical :
prior to lifting the reactor head to ensure that the rated capacity of' the polar crane would not be exceeded during the lift. l The inspectors observed the preparation of_ the reactor head, removal of L
the miscellaneous items, and installation of the lifting device
[ (tripod). During the reactor head lifting,'the inspectors observed that the head was effectively liftod and lowered onto three cribbings on the ;
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. floor to support the reactor head. The maximum load cell reading 'during the lift was 255,570.lbs. The inspector determined that the preparation !
and lifting of the reactor head was acceptable and met all' requirements.
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3.2 Observation of Valve Packing and Modification (62700) -
The inspectors observed portions of valve repacking and modification.
. activities to verify that the maintenance and modification activities were performed in~accordance with the applicable procedures and work orders. The procedure used was the M-0043, Rev.17, Valve Packing. The inspectors observed portions of the following valve maintenance or modification activities.
Valve Valve W/0 or Procedure
.No. Function Location Used- Activities 1403 Isolation ~ Pressurizer M-0043 Packing 1405 Isolation Pressurizer M-0043 Packing 1200 Safety Pressurizer PWO 61/3604 Modification MV-02-2 Isolation Charging Pump -M-0043 Packing Discharge Safety Relief' Valve 1200 had previously been identified to be leaking.
The repair included replacement of the valve stem, but the licensee could not procure an identical stem. Therefore, the l'icensee enlarged the valve stem hole for a larger stem. The modification process and requirement for a, liquid penetrant examination were stated in work order PWO 61/3604. The inspectors determined that all the valve repackings and the modification stated above were performed in an acceptable manner.
3.3 Document Review of Main Steam Safety Valve Setpressure Testing (62700)
The licensee completed setpressure testing for.16 Main Steam safety valves for Unit I during the early part of the refueling outage. The procedures used for the testing and calibration of measuring and testing equipment (M&TE) were: '
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GMP M-0705, Rev. 27, Main Steam Safety Value Maintenance and 4
Setpressure Testing Q1-12-PR/PSL-1, Rev. 21, Calibration of Measuring and Testing l Equipment 4
Q1-12-PR/PSL-2, Rev. 20, Control and Calibration of Measuring and i Test Equipment (M&TE) l l The inspectors reviewed 16 packages of testing records and found that all of them met the requirements of Procedure GMP M-0705 and were acceptable. The 16 valves were divided into train "A" and "B" as shown
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5 Train A Train B Valve No. 8201, 8202, 8203, 8204 8205, 8206, 8207, 8208 -
t 8209, 8210, 8211,-8212 8213, 8214, 8215, 8216 i .
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4 i The required records and data for the setpressure testing are contained in Appendix B, Determining Safety Valve Setpressure with Air Setpressure Device, of GMP M-0705.
. Ten valves, 8203 to 8208 and 8213 to'8216 as listed in Condition Repo'rts l (CR)96-597 were found to have a setpressure outside the acceptable boundaries during the first test and required evaluation for tolerance change based on fuel analysis. The acceptable ranges were defined as
+/- 1% of midpoints 985 psig or 1025 psig.- 1 The M&TE used for testing the main steam safety valves included a specific pneumatic motor and several pressure gauges. The calibration for the motor and most of the gauges in the pre-calibration occurred one l week before work started and in the post-calibration one or two days . i after the work was completed. The exception was the pre-calibration for J 4
gauges M-222 and M-288 six months prior to use. The inspectors !
questioned the licensee as to the pre-calibration validity for these two l gauges. The licensee explained that the pressure gauges had not been 1 used since the last calibration and the effective calibration date had )
not expired. Thus the previous calibration was still valid.and could be 3 used as a pre-calibration. The inspectors concurred in the licensee's !
explanation and verified that all the gauges had post-calibration !
immediately after the tests were completed and that all the gauges were l found to be within the allowable range and acceptable. .The inspectors also reviewed the M&TE checkout log and history for the pressure gauges used in the testing and identified that gauges M-222 and M-288 had not been used since October 15, 1995 when they were post-calibrated. The l inspectors concluded that the licensee performed adequate s'etpressure l tests for the main steam safety valves and used validated pressure gauges for this test.
D. 7.0 Exit The inspection scope and results were summarized by the NRC Senior Resident Inspector with his regular monthly exit on May 13, 1996. The l inspectors described'the areas inspected and discussed in detail the l inspection results. Proprietary information is not contained in this report.
, E. 8.0 Review of FSAR Commitments Note: Please insert the Standard Paragraph with no different findings.
. F. 9.0 Acronyms
- ANSI - American National Standard Institute CE - Civil Engineering GMP - General Maintenance Procedure 2 -1 l
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5 G. Summary Statement:
The procedures used for the testing and surveillance activities reviewed, and also those for the modifications and maintenance activities reviewed were adequate to provide the details for craft personnel to conduct work (paragraphs 3.1, 3.2, and 3.3). Craft i personnel were knowledgeable and skillful in implementing the procedures for those activities observed during the inspection.
G. IFS Forms: None H. Completed NOV: NONE I. Cover Letter Paragraph: NONE ,
J. NOTE FOR DRS ONLY: The report file name is S:\DRS\ SIB \STL9606.RCC f
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86-dB-1996.05:36PM St Luc 1e Re: int'044es 1 4!r/ 461 4622 P.02 ;
l ST. LUCIE UNIT 1
'HIDLOOP/ REDUCED INVENTORY ACTIVITIES JUNE 18, 1996 l On June 18.1996. the licensee plans to enter reduced inventory for approximately l
two days to replace the pump seals for Reactor Coolant Pum) 1A2 The current '
. schedule is to start reduction in the RCS inventory between :.2:00 and 5:00 a.m.
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on June 19.1996. In accordance with Region II Office Instruction 2216. a review of the following items were verified priqr to this evolution:
- Containment Closure Capability - Instructions were issued . to accomplish containment' closure. The equipment hatch was op'en but the licensee plans to close this penetration prior to being the reduction in RCS inventory. The inspector reviewed the penetrations which are to remain open at the time of the drain down and verified that. closure capability was available,
e RCS Level' Indication - Independent RCS wide and narrow range level instruments, which indicated in the controi room, were operable. An additional Tygon tube loop level in the containment was installed and was visible to a dedicated operator in the control room via a j television monitor.
e RCS Level Perturbations - When RCS level reduction is initiated.
additional operational controls will be invoked. Operations will
- take appropriate action to ensure that maintenance will not perfom work that might effect RCS level or shut down cooling.
4 e RCS Inventory Volume Addition Capability - The HPSI peps. and the B !
- charging pump were available for. inventory addition, as were two l trains of shutdown cooling.
i e 'RCS Nozzle Dams - The RCS nozzle dams previously installed in each '
- . of the two steam generators will be inspected for integrity every l l four hours.
e Vital Electrical Bus Availability - Operations does not plan to i release busses or alternate power sources for work while the unit is in a reduced inventory. l e Pressurizer Vent Path - The manway atop the pressurizer has been .
removed to provide a vent path. Operations will verify that the !
r manway is' unobstructed every four hours, j
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