ML17216A412
| ML17216A412 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/23/1986 |
| From: | Woody C FLORIDA POWER & LIGHT CO. |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR 10CFR50.61, L-86-20, NUDOCS 8601290123 | |
| Download: ML17216A412 (13) | |
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, tl REGULATORY ORMATION DISTRIBUTION BY M (RIDS)
ACCESUSISN-NBR+860ISBDI43 DOC;DATE~6801-$33NOTARIZED'NO.
DOCKET-S~
FACIL: 50-335 St.
Lucis Plant, Unit 1
Florida Poasr a8 Light Co.
05000335 AUTH. NAME AUTHOR AFFILIATION MOODYP C. O.
Florida Power 5 Light Co.
RECIP. NAME RECIPIENT AFFIL1ATION MIRAGLIA>F. J.
Division oF Pressurized Mater Reactor Licensing B (post 8
SUBJECT:
Notifies that max RT (pressurized thermal shack) of 230 F will be experienced bg lower shell longitudinal seams at end of QLd per 10CFR50. 61 (b ) ( 1 ). Value consistent w/screening criterion for long itudinal we 1 d s.
s DISTRIBUTION CODE:
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TITLE:
OR Submittal:
Thermal Shock to Reactor Vessel NOTES:
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FLORIDAPOWER & LIGHTCOMPANY Mr. Frank J. Miraglia, Director Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C.
20555 JAN 2 3 1886 L-86-20
Dear Mr. Miraglia:
Re:
St. Lucie Unit 1 Docket No 50-335 10 CFR 50.61 (b)(1) Re ort 10 CFR 50.61 (b)(l) requires the submittal of projected values of RT (PTS) for the end of the operating license for St. Lucie Unit l.
This requirement arises from the concern over reactor vessel embrittlement and pressurized thermal shock.
As part of the FPL effort associated with the St. Lucie Unit 1 thermal shield removal in 1983, we have previously submitted the majority of the required data. At that time, it was concluded that St. Lucie Unit 1 could safely operate without restriction from PTS considerations.
Since that time we have refined our fluence predictions and analyzed a surveillance capsule. It is our determination therefore, that a maximum RT (PTS) of 230OF will be experienced by the lower shell longitudinal seams at the termination of the plant operating license (July, 2010).
This value compares favorably to the screening criterion of 270OF for longitudinal welds.
Should you or your staff have any questions on this information, please contact us.
Very truly yours, C. O. Woo Group V'esident Nuclear Energy Department SAC/cac Attachment cc:
Dr. J. Nelson Grace - Region II, USNRC Harold F. Reis, Esquire PNS-Ll-86-10 60 lPP0123 Og000335 g60 $23 pDR ISIDOCW 0 Png P
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ATTACHMENT1 Reactor Vessel Beltline Materials Descri tion The St. Lucie Unit 1 reactor pressure vessel was designed and fabricated by'ombustion Engineering, Inc.
The reactor vessel beltline, as defined by 10 CFR 50, Appendix H, consists of the six plates used to form the lower and intermediate shell courses in the
The plates were manufactured from SA533 Grade B
Class 1
quenched and tempered plate.
The heat treatment consisted of austenization at 1600 + 50F for four hours, water quenching and tempering at 1225 +
25F for four hours.
The ASME Code qualification test plates were stress relieved at 1150 + 25F for forty hours, and furnace cooled to 600F.
The longitudinal and girth seam welds were fabricated using E8018-C3 manual arc electrodes and Mil B-0 submerged arc weld wire with Linde 120, 0091 and 1'092 flux.
The post weld heat treatment consisted of a forty hour 1150 + 25F stress relief heat treatment followed by furnace cooling to 600F.
The beltline materials are identified in Tables 1 and 2.
Copper, Nickel and initial RT NDT are also listed in these tables.
Calculations Calculations for present and end of license RT (PTS) are based on equation 1 of 10 CFR 50.61 (b)(2) which gave a lower value than equation 2. Allmaterials data used for these calculations are given in Tables 1 and 2.
Source material for these data is supplied in the references.
Generic values for "I" are used for all materials with the associated 59oF "M" value except for a lower shell plate identified as heat number C-5935-3 and the intermediate to lower girth seam where actual values were used with'the associated 08oF "M"value Fluences are projected to end of license based on cycle specific predicted fluence and extrapolated to the end of license based on calculated power distributions.
The models used to predict fluence have been reported in reference 0.
TABLE 1 REACTOR VESSEL BELTLINEPLATES ST. LUCIE UNIT 1 LOCATION HEAT NO-SUPPLIER 9fi Cu<
% Ni<
RT(NUT) U>
Intermediate Shell Intermediate Shell Intermediate Shell Lower Shell Lower Shell
, Lower Shell A-0567-1 B-9027-1 A-0567-2 C-5935-1 C-5935-2 C-5935-3 Lukens Lukens Lukens Lukens Lukens Lukens 0.11 0.11 0.11 0.15 0.15 0.12 0.60 0.60 0.58 0.56
. 0.57
, 0.58 OoF OoF 10oF 20oF 20oF~3>
OoF
C t
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TABLE2 REACTOR VESSEL BELTLINEWELDS ST. LUCIE UNIT 1
.Location i
Weld Sean No.
Wire Heat No.
Flux'~
Flux Batch
% Cu~ I) % NI ~ I)
IGNIS'a~ I)
Interrrediate Shell Longi-tudinal Sean Internmdiate Shell Longi-tudinal Seam Intermediate Shell Longi-tudinal Sean 2-203A 2-203B 2-203C A-8706/30B009 Linde 120 3878 R 3688 0.12 0.20
-56 M/A IPQI+
A-8706/30B009 Linde 120 3878 R 3688 0.12 0.20
-56 M/A IPCI+
M/A CBB+
A-8706/30B009 Linde 120 3878 R 3668 0.12 0.20
-56 Lower Shell Longitudinal Sean 3-203A 30B009 M/A KBE3+
Linde 1092 3889 0.30 0.60
-56 Lawer Shell Longitudinal Sean 3-203B 30B009 M/A KBE3+
Linde 1092 3889 0.30 0.60
-56 Lawer Shell 3-203C Longitudinal Seam
- 30B009 Linde 1092 3889 0.30 0.60
-56 Intermediate to Lower Girth Seam 9-203 90136 M/A ABEA+
M/A FCAA+
Linde 0091 3999 0.23 O.ll
-60~3) 4/lanual shielded rretal arc electrode (all others autonatic suhn rged arc wire).
TABLE 3 FLUENCE AND RTPTS 1986 and 2010 Location Interrradiate Shell Interrn diate Shell Intermediate Shell
. Laver Shell Laver Shell Laver Shell Interrrediate
. Shell L'ongi-tudinal Seam Intermediate Shell Longi-tudinal Seam Intermediate Shell Longi-tudinal Seam Lawer shell Longitudinal Seam Lever Shell Longitudinal Seam Laver Shell Longitudinal Seam Intermediate to Lcnver Girth Seam Fluence n/cm 3an 1986 3.39 x 1018 3.39 x 1018 3.39 x 10i8 3.39 x 1018 3.39 x 1018 3.39 x 1018 3.39 x 1018 3 39 x 1018 3.39 x 1018 3.39 x 1018 3.39 x 1018 3.39 x 1018 3.39 x 1018 RT PTS 38Ã 1986 109 109 117 106 136 112 151 151 151 Fluence n/on XLY 2010 1.68 x 1019 1.68 x 1019 1.68 x 1019 1.68 x 1019 1.68 x 1019 1.68 x 1019 1.68 x 1019 i.68 x io>>
1.68 x io>>
1.68 x 1019 i.'68 x i019 1.68 x 1019 i.68 x i019 RT PTS op XLY 201(
135 135 103 182 172 100 66 66 66 "230 230
- 230
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References 1.
FPL Letter, L-83-263, St. Lucie Unit 1 Reactor Vessel Internals and Therrral Shield, Plant Recovery Progran, R. E. Uhri g to R. A.
Clark, April 27, 1983 2.
Ccrrhustion Engineering Report, F-NLM-007 Program For Irradiation Surveillance of Hutchinson Island Plant
{PSL-.I) Reactor Vessel Materials, September 15, 1970 3.
Ccrrhustion Engineering Report, IR-F-MM-005 Evaluation of Baseline Specimens, FPL St. Lucie Unit 1
0.
FPL Letter, L-80-29, Reactor Vessel Internals and Thermal Shield Conclusions and Findings, St. Lucie Unit 1, 3. W. Willians, 3r.
to 3. Miller, February 10, 1980
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