ML17213B294
| ML17213B294 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/15/1983 |
| From: | Sells D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TAC-51343, NUDOCS 8304260019 | |
| Download: ML17213B294 (23) | |
Text
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APR 1 5 t983 gg~>~~
yg PQ-Olb Docket No. 50-335 LICENSEE:
Florida Power II Light Company FACILITY:
St. Lucie Plant Unit No.
1
SUBJECT:
THERIIAL SHIELD INVESTIGATION IhR IhPJA
, HPJL
'Ko HO IIA O
OH oo NA Ol<
OIL P)A Le
'FFiCEI SURNAME/
DATE Q On April 12, 1983 a meeting was held in Bethesda, IId. to provide the NRC staff with the current status and future plans of the thermal shield investigation being conducted by Florida Power and Light (FP8L) at the St. Lucie Plant, Unit 1.
The NSSS supplier (Combustion Engineering CE) and the licensees for Fort Calhoun, IIaine Yankee and Hillstone 2 also attended the meeting (Attachment 1 is a list of attendees),
Combustion Engineering stated that a Part 21 review was conducted.
Since they believe that the worst case condition would be the case where the thermal shield drops, they considered the results of such a failure.
CE believes that the snubbers near the bottom of the core support barrel would arrest the falling shield and have no significant impact on flow through the core.
Since the snubbers were not previously analyzed to consider this
CE stated 1
en act to arrest the fall.
The core sto s are FP8L opened the meeting with a presentation that covered a brief chRonology of events leading to the discovery of loose parts in the core area and a current status report on actions taken to date and proposed in the future.
The damage identified to date is all related to the thermal shield.
This includes damage to the thermal shield support lugs (welded to the core support barrel), thermal shield support pins, upper positioning pins and locking
- bars, lower positioning pins and locking bars, the thermal shield and evidence of damage to the core support barrel.
Attachment 2 includes viewgraph slides giving the preliminary results of the initial inspection of the thermal shield and its supports and the vessel/internals/fuel interfaces.
The only damage detected to date is associated with the thermal shield.
No damage to the core (i.e., fuel) has been found.
St. Lucie 1 has a loose parts monitoring system.
Signals have been observed in the past, but no safety significance was attached to these signals by FPSL.
These signals have been intermittent since 1978.
Viewgraph 8 of Attachment 2 lists the current action items that will be under-taken by FPEL.
No schedule for completion was developed at the time of the meeting.
Plans are not yet complete, but it is anticipated that one surveillance capsule will be removed for analysis and a more comprehensive ISI program will be pursued than originally planned for this outage.
The conclusions of the investigation are listed on Viewgraph 9 of Attachment 2.
The bottom line is that problems resulting from the damage did not endanger the core or reactor vessel since no abnormal flow occurred that could have caused a significant safety problem.
No failure mode has been established as yet and a full schedule has not been developed.
It is believed that failure was not a single event.
The results of the investigation and their evaluation will be the sub-ject of a future presentation.
NRC FORM 318 (10-80) NRCM 0240 OFF I CIA L R ECOR D COPY US0 PO: 1981-335-960
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w 2 designed to stop the fall of the core support barrel assembly with the in-place core and attached thermal shield. 1f the support flanges should fail.
Again, no significant impact on flow through the core would result.
CE also indicated that if damage to the fuel should occur, then other instru-mentation would detect such damage prior to the occurrence of a significant failure.
Since the thermal shield was in no danger of falling, flow blockage would not occur even if the thermal shield were to fall, and back up instru-mentation would give early detection of damage, CE stated that it was their conclusion that there was no reason to recommend a shutdown for the other CE plants that have an installed thermal shield.
CE believes that damage at St. Lucie 1 occurred over a period of time and that it will take a month or longer to complete its failure analysis.
They did express a willingness and a desire to meet with the staff as results became available from their analysis to keep the staff advised of the progress being made.
The other three licensees with CE designed plants conta1ning thermal shields made brief presentations.
Viewgraphs used by the Fort Calhoun licensee are provided in Attachment 3.
A complete 10 year Inspection was recently per-formed, with particular emphasis on inspection of the thermal shield.
IIo evidence of damage was found and the plant has been returned to power.
I<aine Yankee found 2 upper positioning pins missing and one damaged during its last refuel1ng outage.
This was discussed with the staff in October 1982 and a report was submitted concerning th1s event in February 1983.
It is believed that relaxation of the pre-load led to buffeting and directly to the problem encountered at Maine Yankee.
The plant is about 20Ã through cycle 7 at present.
Spring 1984 is the next scheduled outage.
Hillstone 2 is scheduled to beg1n its next refueling outage in approximately seven weeks.
There have been no anomalies that would indicate that any problems exist within the core area.
Nllstone 2 most closely resembles St. Lucie 1 in des1gn.
After a short staff caucus the staff requested that FP&L and the other licensees take certain actions.
The actions requested of FP&L include:
1.
Continue to provide close communications with the staff on the status of the investigation.
2.
Document in a future submittal(s) the results of the inves<<
- tigation, a failure analys1s, plan for recovery, and any safety assessments needed to either change reactor configuration or operation.
This should include consideration of vessel 1n-teraction, broadened scope to include full exposure.
vessel exposure and integrity of core internals, to name a few..
3.
Commit to provide the above information and documentation to obtain staff review and concurrence prior to restart.
OFFICE%
SURNAMEII DATE P
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Licensees of the other plants were asked to:
l.
Document the presentation provided during the meeting.
2.
Commit to establishing a firm interface with the St. Lucie 1 program.
3.
Provide justification for continued operation to in-clude at a minimum, an analysis similar to CE's Part 21 review.
4.
Provide a description and operational experience of the loose parts monitoring system installed at the plant. If validity of the system is questioned, pro-vide a description of back-up measurements that are considered and used, such as, ex-core instrumentation and axial power distribution measurements.
5.
Develop a plan and schedule that will utilize the St. Lucie 1 results as they apply to the specific plants.
6.
Be prepared to respond to requests for additional information such as reports on loose parts monitoring systems and opera-tions and logs from other instrumentation.
Each licensee was asked to develop the above inputs on the bases of the re-quests made at the close of the meeting.
Licensees other than FP&L were requested to submit the above requested information by April 26. 1983.
Gr7gL'nial bTone,'d Sp Attachments:
As Stated cc:
See next page Donald E. Sells Pro)ect tanager Operating Reactors Branch P3 Division of Licensing OFFICEI SURNAME/
DATE I}
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NRC FOAM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USQPO: 1881~5.880
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MEETING
SUMMARY
DISTRIBUTION Licensee:
Florida Power and L'ight Company
- Copies also sent to those people on service (cc) list for subject plant(s).
Docket File NRC PDR L PDR NSIC ORB83 Rdg ORB83 Summary File JHeltemes BGrimes RAClark Project Manager PMKreutzer OQ D ELJordan JMTaylor ACRS-10 NRC Participants J.
Ridgely-ASB V. Nerses-LB 83 MS128 A. Wang ACRS-MS 1016 H St.
Mr. Jeffrey Soper P. 0.
Box 3434 Fort Pierce, Florida 33450
0
Attachment 1
'IST Of ATTENDEES THERMAL SHIELD INVESTIGATION ST.
LUCIE PLANT, UNIT 1 APRIL 12, 1983 NRC G; Lainas F. Miraglia W. Johnston R. Clark T. Ippolito D. Sells L. Phillips M. Tokar G.
Schwenk L. Lois D. Smith W. Hazelton R. Kiessel C.
Chen/'.
Tourigny K. Heitner D. Fieno H.
Shaw P.
Leech J. Blake, Reg. II L. Yandell, Reg.
IV ED Kelly, Reg.
I FPSL D. Chancy T. Dillard D.
James R.
Dawson J.
Zudans (NUS)
Northeast Utilities R. Laudenat S.
Chandra M. Cass P. Parulis OPPD T. Patterson T. McIvor R. Shimkus (SRI)
'CE A. Scherer F. Grubblich S. Ritterbusch R. Jacques T. Gates R.
Longo C. Brinkman J.
Kingseed Maine Yankee J. Garrity C.
Eames
Attachment 2
AGENDA MEETING ON ST.
LUCIE UNIT 1
THERMAL SHIELD
- TUESDAY, APR IL 12, 1983 1:00 p.m.
MARYLAND NATIONAL BANK BUILDING, ROOM 6507
- BETHESDA, MD 1:00 INTRODUCTIONS AND PURPOSE OF MEETING 1: 15 CHRONOLOGY LEADING TO PRESENT.
SITUATION 1: 30 REACTOR
'lESSEL INTERNALS DESCRIPTION AND DAMAGE
SUMMARY
1:45 V IDEO TAPE PRESENTATION 2:00 ACTION ITEM
SUMMARY
AND SCHEDULE FOR COM-MUNICATIONSS MITH NRC
PRESENTATION OUTLINE OF INSPECTION RESULTS
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DESCRIPTION OF THE ARRANGEMENT OF THE INTERt<ALS IN THE REACTOR VESSEL.
DESCRIPTION OF THE MAJOR INTERNALS COMPONENTS.
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DESCRIPTION OF THE THERMAL SHIELD AND SUPPORTS.
REPORT OF THE INSPECTIOt<
RESULTS.
A.
VESSEL -
INTERNALS -
FUEL INTERFACE B.
THERMAL SHIELD AND SUPPORTS INTERFACES
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VIDEO TAPE REYIEll.
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IN-CORE INSTR UMENTATION ASSEMBLY IN-CORE INSTR,UMENT GUIDE TUBE CONTROL.
ELEMENTS SEMBLY.
FULLY W ITHDRAWN CEDM NOZZLE I NSTR UMENTATI ON NOZZLE ALI G NMENT
- PIN, UPPER GUIDE STRUCTURE 42" ID OUTLET NOZZLE S UR VEILLANCE HQUlER 136. 7" ACTIVE CORE LENGTH 3O Il x INLET NOZZLE CORE SUPPORT BARREL CORE SHROUD FUEL ASSEMBLY THERMAL SHIELD SNUBBER CORE STOP
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CORE SUPPORT ASSEMBLY FLOW SKIRT FLORIDA POWER 8 LIGHT CO.
St.
Luci.e Plant Reactor Arrangement - Vertical Section I1 Figure
- 4. L-L
Expansion Compensating Ring 0
0 o
Upper Guide Structure Support Plate Alignment Key EA Shroud Outlet Nozzle Jn>>Core iiistrumenta tion Guide Tube Alignment Pins I
Fuel Alignment Plate Core Support Barrel Core Shroud C
Thermal Shield Fuel Alignment Pins
'Core Support Plate Core Support Assembly Snubber FLORIDA POWER 8 LIGHT CO.
St.
Lucie Plant Unit 1 Reactor Internals Assembly Rev.
13 - 7/15/7:-
Fiovr
- 4. 2-
TIIERINL SHIELD SUPPORT PLUG TIIENSL SIIIELD i
I CORE SUPPORT BARREL TIIENCL SIIIELD SUPPORT LUG
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.TIIENIAL SIIIELO SUPPORT
'IIIH TIIERIIAL SIIIELO SUPPORT SYSTEtl LOCK BAR POSITIO/IIHG PIN
VESSEL -
INTERNALS.-
FUEL 'INTERFACES INSPECTION RESULTS
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CORE BARREL.AND REACTOR VESSEL SNUBBER SURFACE APPEAR NORMAL.
CORE BARREL -
UPPER GUIDE STRUCTURE ALIGNMENT KEYS AHD MATING SLOT SURFACE APPEAR NORMAL.
FUEL ALIGNMENT PLATE SLOTS AND MATING CORE SHROUD GUIDE LUG "CONTACT SURFACES APPEAR NORMAL.
CORE PLATE FUEL ALIGNMENT PINS APPEAR NORMAL.
IH
SUMMARY
, THERE IS HO SIGNIFICANT WEAR OR RELATIVE MOTION BETWEEN ANY INTERFACES.
OVERALL
SUMMARY
OF INSPECTION RESULTS THERMAL SHIELD AND SUPPORTS TWO TAPER SUPPORT'LUGS MISSING WITH PIECES OF RIM OF THERMAL SHIELD. C,lt0~~ Sto < ~~~)
TWO UPPER POSITIONING PINS MISSING. (>>o ~ 3/g 4
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FOUR THERMAL SHIELD SUPPORT LUGS HAVE PULLED AWAY FROM THE BARREL. ( 3o ~vo jfqg~ >>o ~~~~~
9 THERMAL SHIELD HAS MOVED DOWN ABOUT 1
TO 2
INCHES.
WEAR ON TOP OF FIVE SUPPORT LUGS.
FOUR LOWER POSITIONING PINS PROTRUDE'ROM THERMAL SHIELD.
SIX LOCK BARS MISSING.
ACTION ITEM LIST 1.
REMOVAL AND EVALUATION OF LOOSE PARTS.
2, IDENTIFICATION AND EVALUATION OF DAt1AGED COMPONENTS, 3.
PREPARATIONS FOR THERMAL SHIELD REMOVAL.
4.
CONSEQUENCES OF THERMAL SHIELD REMOVAL:
EVALUATION AND REPAIR OF INTERNALS ANALYSIS OF MARGIN EFFECTS 5.
PREPARATION FOR INSERVICE INSPECTION.
6.
PREPARATION FOR SURVEILLANCE CAPSULE REMOVAL.
7.
ANALYSIS OF FAILURE MECHANISM.
CONCLUSIONS 1.
NO SAFETY PROBLEM WAS ASSOCIATED WITH THE FAILURE.
A.
THERMAL SHIELD (T.S.)
REMAINED SUSPENDED.
B.
IF T.S.
HAD DROPPED, NO DISRUPTION TO CORE FLOW (SNUBBERS/CORE STOPS WOULD RETAIN) 2.
DEBRIS DID NOT SIGNIFICANTLY AFFECT COOLANT FLOW.
3.
'/I SUAL INSPECTION INDICATES NO ABNORMAL REACTOR INTERNALS V I B RAT I ON.
Attachment 3
Fort Calhoun Core Support Barrel and Thermal Shield Inspection I.
Original Plan - Exterior of Core Support Barrel A.
Perform visual inspection of al 1 accessible attachment welds per ASME Section XI, Table IWB-2S00.
B.
Inspect points of attachment (snubbers) of Core= Support Barrel to Reactor Vessel.
C.
Generalized inspection of exterior of Core Support Barrel and Thermal Shield to determine general condition after 10 years of service.
II.
Inspection was performed by the District's ISI contractor - South-west Research Institute III.
Oue to the concerns expressed by CE AOP Infobulletin 82-12, it was decided to emphasize the inspection of the heads of the thermal shield positioning pins, the locking collars, and the lock welds.
.0 IV.
Oifferences between Fort Calhoun and other CE reactor vessels with thermal shields A.
Size - Fort Calhoun vessel is smaller l.
8 lugs to support Thermal Shield from Core Support Barrel.
2.
8 upper positioning pins below support lugs.
3.
16 lower positioning pins between support lugs.
4.
Other vessels have 9 lugs, 9 upper pins, 17 lower pins.
B.
Positioning pin locking attachment 1.
Fort Calhoun - locking collar welded to both pin and Thermal Shield.
2.
Others - locking bar across head of pin, welded to Thermal Shield only.
C.
Annulus between Thermal Shield and Core Support Barrel 1.
Fort Calhoun 3/4" 2,
Others 5/8" 3.
Narrow annulus rendered TV camera inspection of that area virtually impossible.
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4 V.
Results of Core Support Barrel/Thermal Shield Inspection A.
All upper and lower positioning pins, locking collars, and lock welds were verified to be intact.
t<o evidence of wear or crack-ing was noted.
B.
All accessible snubbers (3 of 6) were examined.
No evidence of wear between the Core Support Barrel and the Reactor Vessel vas noted.
All 6-supports in the Reactor Vessel were examined, with no deterioration found.
C.
All accessible Thermal Shield support lugs (4 of 8) were exam-ined, with no evidence of deterioration.
D.
Accessible portion (more than 180') of the Core Support Barrel upper to lower shell weld was
- examined, with no indications found.
E.
Accessible portion (more than 180') of the Core Support Barrel to flange weld was examined.
Ho indications were noted.
F.
One outlet (hot leg) nozzle was examined, with no indications noted.
VI.
Conclusion - Fort Calhoun Core Support Barrel and Thermal Shield are in excellent condition.