ML17213B314

Summarizes 830425 Meeting W/Nrc Re Reactor Vessel Internals & Thermal Shield Plant Recovery Program Concerning Pressurized Thermal Shock.Plant May Safely Operate W/O Restriction from Pressurized Thermal Shock Considerations

ML17213B314
Person / Time
Site:	Saint Lucie
Issue date:	04/27/1983
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR L-83-263, NUDOCS 8305020345
Download: ML17213B314 (23)

Text

REGULA~ Y INFORMATION DI STRIBUTIO YSTEM =('R IDS)

ACCESSION NBR:8305020345 DOC DATE: 83/04/27 NOTARIZED; NO FACIL:50-335 St, Lucie Planti Unitt AD AUTH, NAME AUTHOR

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AFFILIATION ii Florida Power 8, Light Co< 05000335 UHRIG'RNE ~ Florida Power 8 Light Co ~

i REC IP, NAME

'perating RECIPIENT AFFILIATION NOTES'OCKET CLARKiR~

SUBJECT:

Summarizes 8

Reactors 830425 meeting w/NRC Branch 3 re,reactor vessel internals thermal shield plant recovery program concerning pressurized thermal shock.'Plant may safely operate,w/o

.restriction from pressurized thermal shock considerations.

DISTRIBUTION CODE: AOAOS TITLE: OR COPIES RECEIVED:LTR Submittal: Thermal 'Shock to Reactor Yessel

- ENCL l SIEE:

REC IP IENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL NRR ORB3 BC 01 7 7 INTERNAL: ELD/HDS2 12 1 0 MURLEYgT 1 1 NRR DIR 1 1 i NRR V I SS ING G04 1 1 NRR/DE/MTEB 1 1 NRR/DHFS D IR 1 1 NRR/DL DIR 1 1 NRR/DL/ORAB 11 1 0 NRR/DSI DIR 1 1 NRR/DSI/RSB 1 1 N DIR 1 1 NRR/DST/GIB 1 1 RE4 FIL 1 1 RES/DET 1 1 A 1 1 RGN2 1 1 EXTERNAL: ACRS 10 6 6 LPDR 03 1 1 NRC PDR 02 1 1 NSIC 06 1 1 NTIS 1 1 tTOTAL NUMBER OF COPIES REQUIRED: LTTR 33 ENCL '31

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FLORIDA POWER & LIGHT COMPANY April 27, I 983 L-83-263 Office of Nuclear Reactor Regulation Attention: Mr. Robert A. Clark, Chief Operating Reactors Branch $/3 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Clark:

Re: St. Lucie Unit I Docket No. 50-335 Reactor Vessel Internals and Thermal Shield; Plant Recover Pro ram In a meeting on April 25, 1983, Florida Power and Light Company provided you with information about the St. Lucie Unit I reactor vessel with respect to pressurized thermal shock (PTS). This letter sum'marizes that meeting and satisfies item D(l) of our letter of April I 9, I 983 (L-83-230).

CEN-I89, Appendix F ("Evaluation of PTS Effects due to Small Break LOCA's with Loss of Feedwater for the St. Lucie I 8 2 Reactor Vessels" December, l98I),

provided an evaluation of the St. Lucie I reactor vessel for PTS effects of certain specific hypothetical plant transients (post TMI 2 Action). The governing axial weld is located in the lower shell course at an azimuthal location where the fluence profile is 47% of the peak fluence. Using the material and fluence data developed in CEN-I89, Appendix F, the residual chemistry for this weld is 0.30 wt. % Cu and .64 wt. % Ni. The resultant end-of-life RTNDT for this weld is predicted to be 2I7OF using the shift prediction method of SECY-82-465 ("NRC Staff Evaluation of PTS",

November 23, I 982).

For the sake of comparison only, even if St. Lucie I operated from beginning-of-life to end-of-life without a thermal shield, the end-of-life RTNDT predicted by the method of SECY-82-465 would still only be 252OF. Since St. Lucie I has operated until now with a thermal shield, the end-of-life RTNDT would be somewhat less than this upper bound value. We understand the NRC staff and their consultants are developing a revision to the RTNDT shift prediction method of SECY-82-465 which would result in a lower EOL RTNDT prediction.

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Page 2 Office of Nuclear Reactor Regulation Mr. Robert A. Clark, Chief These evaluations will be updated as the results of the surveillance capsule program become available. Also, the vessel beltline weld in-service inspections being completed this outage will provide additional confirmation of the soundness of the vessel material. Based on these evaluations, St. Lucie I may safely operate without restriction from PTS considerations.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems & Technology REU jDAC/cab cc: Harold F. Reis, Esquire Enclosure

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ST, LUCIE -'1 PTS MEETING AGENDA APRIL 25, 1983 1,. REVIEW PRESENT PTS .STATUS

2. EFFECTS OF THERMAL SHIELD REMOVAL ON PTS 3, ONGOING PTS EFFORT E4, SCHEDULES

ST, LUCIE j.

PRESENT PTS STATUS l .. MELD CHENI.STRY .

2, FLUENCE CALCULATIONS 3, RT-NDT PREDICTIONS

FIGURE F6-1 ST. LUCIE El REACTOR PRESSURE VESSEL HAP INITIAL RTNPT IN 'F 20 C ~

6 ~ 3 E3D I 203@@ C. 6-2 Egg";. C 6 II%03

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TABLE F6-1 ST. LUCIE UNIT Nl REACTOR VESSL'L HATERIALS Product Via ter i a 1 Orop Weight Initial ChemicaI Content~i)

Form Identi fi cation ~NDTT 'F RTND~T'F Nic el Phos horus b

Plate C6-1 10 10 .53 .14b '.012 Plate C6-2 -30 -30 .53 .14b . 011 Plate C6-3 -10 -10 .53 .14 .011 Plate C7-1 0 Oa .64 ,11 .004 a

Plate C7-2 -30 0 .64 .11 .004 Plate C7-3 -30 10 .58 .11 .004 Plate CS-1 . -10 20 .56 .15 .006 a

Plate CS-2 0 0 .57 .15 .006 Plate CB-3 0 'c .SSd .12 .004 Wel d 1-203 A,B,8C N/A -50 .

~ 22 .015 Wel d N/A -50 .018 Wel d 3-~03 A,B,tlC N/A

'-50 ~l .013 rf Meld N/A -60 .21 .016 9-203 N/A -60 .11 .23 .013 N/A Not Available a Oetermined using Branch Technical Position HTEB 5-2 b E'stimated based on average of similar plates c Estimated {see Table F6- 2) d Estimated Ni content {high nickel type wire) e Estimated Ni content low nickel type wire)

a ST. LUCIE 1 CURRENT FLUENCE CALCULATIONS (REF, CEN-j.89)

NS-2 . R5 DOT PROFILE CNS-2 POMER DIST.)

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VIS-2 CAPSULE BENCHNARK SL-1 POINT KERNEL PEAK (SL-j. POWER DIST.)

ST. LUCIEI AZIMUTHAL FLUENCE VARIATION AI VESK Ojio INTSFAK HORIIAL)ZED FAST FLUX I.b b.7 8.6, b.k Jh 15

%TA IKNEES FIGURE F5-3

FIGURE F6-2 ST. LUCIE fl REACTOR PRESSURE VESSEL NP ADJUSTEO RTNDT IN 'F (12/31/81) 3.55 Effect)ve Ful 1 Power Years 0.33 x 1019n/cm2 Peak Surface Fluence 20 th ~ I I.203- C ~

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ST, LUCIE - 1 CURRENT STATUS I

EOL PEAK FLUENCE = 2..83 x 10 x/cz 'Ol NEv)

EOL HELD FLUENCE = A7..5X x PEAK 1,,305 x j019 v/c~2 EOL HELD RTNDT = 2 17,.7 F

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ST, LUCIE j.

EFFECTS OF THERMAL 'SHIELD REMOVAL ON PTS 1). EOL FI UENCE

2) EOL RT-NDT

ST. LUCIE 1

"(32 .EFPY M/0 T,S.)

THERMAL SHIELD WORTH~1.75 x FLUENCE EQL PEAK FLUENCE =.+ Q,95 x 10 N/CN EOL WELD FLUENCE + 2.35 x 10 w/cm I

EOL WELD RTNDT > 252,7 F (CONSERVATIVE UPPER BOUND)

ST, LUCI E - 1 ONSOINj PTS EFFORT l) SURVEILLENCE CAPSULE EVALUATION

2) UPDATED FLUENCE CALCULATIONS

5) 1OOX BELTLINE HELD ISI

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Reactor Vessel Vs'essel 263 Core.

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Thermal Shield Thermal Shield Core Support Reactor Barrel Vessel Elevation p~,d nlan Vit View FIGURE F5-4

ST, LUCIE 1 UPDATED FLUENCE CALCULATIONS SL-1 9'DOT (SL-1 POWER DIST, CY 1-5)

SL-1 Hk DOT (SL-1 CY6 POWER DIST.)

SL-1 SURVEILLENCE CAPSULE BENCH1ARK

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