ML17227A497
| ML17227A497 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 07/01/1992 |
| From: | Bohlke W FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-92-01, GL-92-1, L-92-189, NUDOCS 9207080119 | |
| Download: ML17227A497 (35) | |
Text
ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEM r, REGULAT .. INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9207080119 DOC.DATE: 92/07/01 NOTARIZED: YES DOCKET FACIT:50-335 St. Lucie Plant, Unit 1, Florida Power & Light Co. 05000335
" 50-389 St. Lucie Plant, Unit 2, Florida Power & Light Co. 05000389 AUTH. NAME AUTHOR AFFILIATION BOHLKE,W.H. Florida Power & Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards response to GL 92-01,Rev 1, "Reactor Vessel Structural Integrity."
DISTRIBUTION CODE: ROOID COPIES RECEIVED:LTR i ENCL Q SIZE:
TITLE: OR Submittal: General Distribution NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 NORRIS i J 2 2 INTERNAL: ACRS 6 6 NRR/DET/ESGB 1 1 NRR/DOEA/OTSB11 1 1 NRR/DST 8E2 1 1 NRR/DST/SELB 7E 1 1 NRR/DST/SICB8H7 1 1 NRR/DST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 OC LF 1 0 OGC/HDS3 1 0 1 1 RES/DSIR/EIB 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 g yp7+g O~8 I
gr Ql Qt~~
E/I ('u 4 NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
gg S TOTAL NUMBER OF COPIES REQUIRED: LTTR &3- ENCL
P.O. Box14000, Juno Beach, FL 33408-0420 L-92-189 10 CFR 50.4 10 CFR 50. 54 (f)
JUL -1 i992 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 RE: St. Lucie Units 1 and 2 Docket No. 50-335 and 50-389 Generic Letter 92-01 Revision 1 Res onse The Florida Power and Light Company (FPL) response to revision 1 of Generic Letter 92-01 "Reactor Vessel Structural Integrity, 10 CFR 50.54(f)" for St. Lucie Units 1 and 2 is attached.
On March 6, 1992 the NRC issued this generic letter as part of an NRC program to evaluate reactor vessel integrity, to ensure licensees are complying with 10 CFR 50.60 and 10 CFR 50.61, and to confirm commitments made in response to Generic Letter 88-11.
The attached information is provided pursuant to the requirements of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f).
Please contact us if there are any questions about this submittal.
truly yours,
'ery W. H. Bohlke Vice President Nuclear Engineering and Licensing WHB/GRM/kw cc: Stewart D. Ebneter, Regional Administrator, Region .II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant DAS/PSL 4724-92 9207080ii9 92070i PDR ADOCK,05000335
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St. Lucie Units 1 and 2 Docket No. 50-335 and 50-389 Generic Letter 92-01 Revision 1 Res onse STATE OF FLORIDA )
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COUNTY OF PALM BEACH )
W. H. Bohlke being first duly sworn, deposes and says:
That he is Vice President, Nuclear Engineering and Licensing for the Nuclear Division of Florida Power & Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.
H. Bohlke STATE OF FLORIDA COUNTY OF Q The me by this W.
and who
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foregoing instrument H. Bohlke, day of who id (did not) take was acknowledged is personally an oath.
known before 19 R2.
to me Q.d Name of No ry Public Notary Public, State of Florida My Comm. Exp. Feb. 18, 1995 My COmmiSSiOn eXpireS BondsdthruPICHARDlns.Agency Commission No. 0 0
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St. Lucie Units 1 and 2 Docket No. 50-335 and 50-389 Generic Letter 92-01 Revision 1 Res onse RESPONSE TO NRC GL 92-01 FOR ST ~ LUCIE UNIT 2 uestion 1 Certain addressees are requested to provide the following information regarding Appendix H to 10 CFR Part 50:
Addressees who do not have a surveillance program meeting ASTM E185-73, 79, or -82 and who do not have an integrated surveillance program approved by the NRC are requested to describe 'actions taken or to be taken to ensure compliance with Appendix H to 10 CFR Part 50.
Addressees who plan to revise the surveillance program to meet Appendix H to 10 CFR Part 50 are requested to indicate when the revised program will be submitted to the NRC for review. If the surveillance program is not to be revised to meet Appendix H. to 10 CFR Part 50, addressees are requested to indicate when they plan to request an exemption from Appendix H to 10 CFR Part 50 under 10 CFR 50.60(b).
The current (1983) version of Appendix H to 10 CFR Part 50 reads:
That part of the surveillance program conducted prior to the first capsule withdrawal must meet the requirements of the edition of the ASTM E185 that is current on the issue date to which- the reactor vessel was purchased.
Later versions of the ASTM standard may be used, but including only those editions through 1982. For each capsule withdrawal after July 26, 1983, the test procedures and reporting ,requirements must meet the requirements of ASTM E185-82 to the extent practical for the configuration of the specimens in the capsule. For each capsule withdrawal prior to July 26, 1983 either the 1973, the 1979, or the 1982 Edition of ASTM E185 may be used.
The St. Lucie Unit 2 reactor vessel was designed to the 1971 Edition through the Summer of 1972 Addenda of the ASME Code. The St. Lucie Unit 2 Reactor Vessel Surveillance Program was designed to ASTM E185-73. In addition, the materials selected for the surveillance program are the most limiting material. Testing of the only capsule removed to date, capsule W-83, was conducted in accordance with E185-82. The St. Lucie Unit 2 surveillance program complies with 10CFR 50, Appendix H. No further action is required.
t uestion 2.a Cer'tain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:
Addressees of plants for which the Charpy upper shelf energy is predicted to be less than 50 foot-pounds at the end of their licenses using the guidance in Paragraph C.1.2 or C.2.2 in Regulatory Guide 1.99, Revision 2, are requested to provide to the NRC the Charpy upper shelf energy predicted for December 16, 1991, and for the end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken pursuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.
Res onse 2.a The St. Lucie Unit 2 reactor vessel was built at a time when the influence of copper content on ir'radiation embrittlement was becoming better defined. Applying these low copper values from the 2St. Lucie Unit 2 reactor vessel beltline materials to the Regulatory Guide 1.99, Revision 2 prediction for Charpy upper shelf energy (USE), does not result in any of the beltline materials falling below 50 ft.-lbs. USE at or before the end of the current operating license. Below is a table showing the predicted end of license (EOL) upper shelf energy at the 1/4 T location for the limiting beltline material. All material property values are as reported in the St. Lucie Unit 2 FSAR<') unless otherwise noted.
MATERIAL INITIAL USE 0 Cu EOL 1/4 T REG GUIDE 1.99 EOL USE ft-Ib g'rsasvcrsc) Hacacc (a/cm') REDUCTION ft-1b Plate M-605-1 105 0.11 1.83 x 10'~ 234 81 Intermediate to Lower Shell Girth 0.07 1.83 x 10 24% 87 Weld (101-171)
'he limiting beltline weld for the St. Lucie Unit 2 is the intermediate to lower shell girth weld (101-171), and is in the St.
Lucie Unit 2 Reactor Vessel Surveillance Program. The limiting beltline plate is the M-605-1 plate and is contained in the St.
Lucie Unit 2 Reactor Vessel Surveillance Program.
uestion 2.b Addressees whose reactor vessels were constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the considerations given to the following
material properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.A of 10 CFR Part 50, Appendix G:
(1) The results from all Charpy and drop weight tests for all 2unirradiated beltline materials, and the unirradiated reference temperature for each beltline material, and the method for determining the unirradiated reference temperature from the Charpy and drop weight test; (2) The heat treatment received by all beltline and surveillance materials; (3) The heat number for each beltline plate or forging and the heat number of wire. and flux lot number used to fabricate each beltline weld; (4) The heat number for each surveillance plate or forging and heat number of wire and flux lot number used to fabricate the surveillance weld; (5) The chemical composition, in particular the weight in percent of copper, nickel, phosphorous, and sulphur for each beltline and surveillance material; and (6) The heat number of the wire used for determining the weld and chemical above.
composition if different than Item (3)
Res onse 2.b The St. Lucie Unit 2 reactor vessel was designed to the 1971 Edition through the Summer of 1972 Addenda of the ASME Section III Code. Although not required to respond to this question because of the vessel design date, all the material properties requested in Question 2.b are reported in the St. Lucie Unit 2 FSAR+ and in the latest 10CFR 50.61 PTS report+.
Addressees are requested to provide the following information regarding commitments made to respond to GL 88-11:
(A) How the embrittlement effects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 525'F. were considered. In particular licensees are requested to describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and on the Charpy upper shelf energy.
(B) How their surveillance results, on the predicted amount of embrittlement were considered.
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(C) If a measured increase in reference temperature exceeds the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Revision 2, or if a measured decrease in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.1.2 in Regulatory Guide 1.99, Revision 2, the licensee is requested to report the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December,16, 1991, and for the end of its current license.
Res onse 3.a The St. Lucie Unit 2 reactor vessel has not operated below 525'F.,
as measured by the cold leg temperature, for a significant period of time. The cumulated time and neutron fluence accumulated at critical operations below 525'F. is conservatively estimated to be less than 8 effective full power hours or significantly below 1 x 10'/cm . Since the resultant fluence is significantly below 1 x 10'/cm, the effect of this low temperature irradiation on the adjusted reference temperature and upper shelf is negligible.
Res onse 3.b Florida Power & Light Co. responded to the NRC Generic Letter 88-11 with L-88-498<4> for both the St. Lucie Units 1 and 2 reactor vessels. That letter stated that the P/T limit curves were in the process of being revised and would use the latest Regulatory Guide 1.99, Revision 2 methods for calculation of adjusted reference temperature. On February 7, 1990@, Florida Power & Light Co.
submitted Amendment No. 46 for St. Lucie Unit 2, revising the P/T limit curves and LTOP setpoint analyses. The calculation of adjusted reference temperature used the prediction methodology in Regulatory Guide 1.99, Revision 2, Position 1.1. This position was used because only one capsule has been pulled from the four capsule program. Two or more credible surveillance data points are needed to base adjusted reference temperature prediction on actual data per Regulatory Guide 1.99, Revision 2, Position 2.1. Actual Surveillance data from St. Lucie Unit 2 has shown that the Regulatory Guide 1.99, Revision 2 does conservatively over predict the shift in adjusted reference temperature for the limiting plate and limiting weld material. Thus the Regulatory Guide prediction of adjusted reference temperature for the St. Lucie Unit 2 beltline material used in the P/T limit and LTOP analysis is conservative and bounding when actual surveillance data is considered. On August 21, 1990, the NRC issued a letter that said the review of the P/T limit curves (Amendment 46), were acceptable, which completed the review of the 88-11 issue for St. Lucie Unit. 2.
I' Res onse 3.c The measured reference temperature for the St. Lucie Unit 2 surveillance capsule material limiting plate and limiting intermediate to lower shell girth weld from capsule W-83+ have both exhibited an increase in adjusted reference temperature which was less than predicted by Regulatory Guide 1.99, Revision 2.
The measured drop in upper shelf energy for the St. Lucie Unit 2 limiting base metal and limiting weld metal from capsule W-83+ was less than predicted by Regulatory Guide 1.99, Revision 2.
1 REFERENCE LIST (1) Florida Power & Light Co., St. Lucie Plant Unit 2, Updated Final Safety Analysis Report, Chapter 5.0 (2) 'nalysis of. Capsule W-83, Florida Power & Light Co., St.
Lucie Plant Unit 2, Babcock & Wilcox, September 1985, BAW-1880 (3) FPL Letter, L-86-25, St. Lucie Unit 2 10CFR 50.61 PTS Report, C. 0. Woody to NRC, January 23, 1986 (4) FPL Letter, L-88-498, St. Lucie Units 1 and 2 Response to Generic Letter 88-11, W. F. Conway to NRC, November 23, 1988 (5) FPL Letter, L-90-46, St. Lucie Unit 2 Proposed License Amendment, P/T Limits and LTOP Analysis, J. H. Goldberg to NRC, February 7, 1990
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St; Lucie Units 1 and 2 Docket No. 50-335 and 50-389 Generic Letter 92-01 Revision 1 Res onse RESPONSE TO NRC GL 92-01 FOR ST ~ LUCIE UNIT 1 uestion 1 Certain addressees are requested to provide the following information regarding Appendix H to 10 CFR Part 50:
Addressees who do not have a surveillance program meeting ASTM E185-73, 79, or -82 and who do not have an integrated surveillance program approved by the NRC are requested to describe actions taken or to be taken to ensure compliance with Appendix H to 10 CFR Part 50.
Addressees who plan to revise the surveillance program to meet Appendix H to 10 CFR Part 50 are requested to indicate when the revised program will be submitted to the NRC for review. If the surveillance program is not to be revised to meet Appendix H to 10 CFR Part 50, addressees are requested to indicate when they plan to request an exemption from Appendix H to 10 CFR Part 50 under 10 CFR 50.60(b).
Res onse 1 The current (1983) version of Appendix H to 10 CFR Part 50 reads:
That part of the surveillance program conducted prior to the first capsule withdrawal must meet the requirements of the edition of the ASTM E185 that is current on the issue date to which the reactor vessel was purchased.
Later versions of the ASTM standard may be used, but including only those editions through 1982. For each capsule withdrawal after July 26, 1983, the test procedures and reporting requirements must meet the requirements of ASTM E185-82 to the extent practical for the configuration of the specimens in the capsule. For each capsule withdrawal prior to July 26, 1983 either the 1973, the 1979, or the 1982 Edition of ASTM E185 may be used.
ASTM E185 was originally issued in 1961 and was revised in 1966, 1970, 1973, 1979 and 1982. Appendix H to 10 CFR Part 50, which was first published in 1973, outlines the requirements for compliance with the pertinent version of ASTM E185. The St. Lucie Unit 1 reactor vessel was designed to the 1965 Edition through the Winter of, 1967 Addenda of the ASME Code. The St. Lucie Unit 1 surveillance program meets ASTM E185-1970, which was the standard in place at the time the surveillance program was designed. This program is described in the St. Lucie Unit 1 FSAR"'. As part of the
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licensing process, the Atomic Energy Commission reviewed the St.
Lucie Unit 1 surveillance program and concluded that the essential material surveillance requirements of ASTM E185-73 and Appendix H of 10 CFR Part 50 were met~). Testing of the surveillance capsule removed after July 26, 1983 was conducted in accordance with E185-
- 82. Based on the favorable Safety Evaluation+, an exemption to
'Appendix H of 10 CFR Part 50 is not necessary.
uestion 2.a Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:
Addressees of plants for which the Charpy upper shelf energy is predicted to be less than 50 foot-pounds at the end of their licenses using the guidance in Paragraph C.l.2 or C.2.2 in Regulatory Guide 1.99, Revision 2, are requested to provide to the NRC the Charpy upper shelf energy predicted for December 16, 1991, and for the end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken pursuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.
Res onse 2.a The prediction of upper shelf energy for reactor vessel beltline limiting weld and plate is very material and plant specific. Using Regulatory Guide 1.99, Revision 2 for predictions, the St. Lucie Unit 1 vessel beltline materials are not predicted to fall below 50 foot-pounds upper shelf energy (USE) at or before the end of the current operating license. Below is a table showing the predicted end of license (EOL) upper shelf energy at the 1/4 T location for the limiting beltline material.
MATERIAL INITIAL USE % Cu EOL 1/4 T REG GUIDE 1.99 EOL USE ft-lb g'ransverse) Hucace (n/cm') 't REDUCTION ft-lb Plate C-5935-2 103') 0.15 2. 01 x 10'~ 284 74.2 (C-8-2)
Lower Shell Long 112<4) 0.30 1.27 x10'~ 444 Welds (3 203 Ag Bg C) 424'2.7 Intermediate to 144>> 0.23 2.01 x10'~ 83.5 Lower Shell Girth Weld (9-203)
- St. Lucie Unit 1 limiting beltline material ** St. Lucie Unit 1 surveillance material
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~ 0 The limiting beltline weld for the St. Lucie Unit 1 vessel is identified as the lower shell longitudinal weld seams. These weld serums were fabricated with the wire heat and flux lot identified in Table 2. Florida Power 6 Light Co. has also identified that this exact CE fabricated weld is contained in the Duquesne Light Beaver Valley Unit 1 Reactor Vessel Surveillance Program<". The unirradiated initial transverse USE data from the Beaver Valley Unit 1 surveillance program is used in the above calculation.
Initial property data for CE fabricated welds is considered to be transferable between vessels since it is a function of material type and fabrication history without influence by reactor designer.
Florida Power Er Light Co., in conjunction with the Combustion Engineering Owners'roup, has submitted a report<@ to the NRC outlining an approach to integrate the irradiated data like that from the Beaver Valley Unit 1 into the St. Lucie Unit 1 Reactor Vessel Surveillance Program.
uestion 2.b Addressees whose reactor vessels were constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the considerations given to the following material properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.A of 10 CFR Part 50, Appendix G:
(1) The results from all Charpy and drop weight tests for all unirradiated beltline materials, and the unirradiated reference temperature for each beltline material, and the method for determining the unirradiated reference temperature from the Charpy and drop weight test; (2) The heat treatment received by all beltline and surveillance materials; (3) The heat number for each beltline plate or forging and the heat number of wire and flux lot number used to fabricate each beltline weld; (4) The heat number for each surveillance plate or forging and heat number of wire and flux lot number used to fabricate the surveillance weld; (5) The chemical composition, in particular the weight in percent of copper, nickel, phosphorous, and sulphur for each beltline and surveillance material; and (6) The heat number of the wire used for determining the weld and chemical above.
composition if different than Item (3)
Res onse 2.b. 1 The St. Lucie Unit 1 vessel was built to ASME Section III 1965 Edition through the Winter of 1967 Addenda. Fracture toughness
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tests were performed on all the beltline materials used to fabricate the St. Lucie Unit 1 vessel. Where testing was not done in"accordance with the Summer 1972 Addenda of the 1971 Edition of ASME III, the NRC Standard Review Plan (NUREG-0800) Branch Technical Position MTEB 5-2 was used to estimate the toughness properties for plate material. Weld properties were based on generic values as listed in the current 10 CFR 50.61 PTS rule.
Table 1 shows the drop weight tests and RT~~ values for the St.
Lucie Unit 1 beltline plate material as well as the method used to determine these values. Table 2 shows the drop weight test and
,RT~~ values for the beltline weld material.
Res onse 2.b. 2 The heat treatment for both the St. Lucie Unit 1 vessel beltline plate, material and surveillance plate material consisted of austenitization at 1600 F. + 25OF. for four hours, water quenched and tempeied at 1225'F. + 25'F. for four hours.
The ASME Code qualifications and surveillance test plates received a subsequent 1150'F. + 25'F. stress relief for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and furnace cooled to 600'F.+. The surveillance weldment received a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> stress relief of 1100'F. to 1150'F.">><'>
Stress relief heat treatment for the vessel beltline materials was 1150'F. + 25'F. The actual time at temperature for the specific beltline welds depends on its fabrication sequence. The intermediate and lower longitudinal welds received a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> stress relief treatment while the. intermediate to lower girth seam received 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> at stress relief temperature<~.
In conclusion, the test materials used to establish beltline fracture toughness properties have been heat treated to be representative of the reactor vessel beltline materials.
Res onse 2.b. 3 The heat number for each beltline plate for the St. Lucie Unit 1 vessel is listed in Table 1. The wire heat number and flux lot for each beltline weld for the St. Lucie Unit 1 vessel is listed in Table 2.
Res onse 2.b. 4 The St. Lucie Unit 1 surveillance plate and weld material are listed in Tables 1 and 2 and are identified by footnote b.
Res onse 2.b. 5 The chemistry values for the St. Lucie Unit 1 vessel beltline materials are listed in Tables 1 and 2.
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Res onse 2.b. 6 Thh chemistry for each beltline plate was taken from the actual plate material. All weld heat chemistries were taken from the actual weld wire heat number with the exception of the nickel value reported for the intermediate shell longitudinal seams in Table 2.
This value was estimated by CE based on a group of high manganese, molybdenum, low nickel, B-4 weld wire<'". No nickel addition was used with this weld.
uestion 3 Addressees are requested to provide the following information regarding commitments made to respond to GL 88-11:
(A) How the embrittlement effects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 525'F. were considered. In particular licensees are requested to describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and on the Charpy upper shelf energy.
(B) How their surveillance results on the predicted amount of embrittlement were considered.
(C) If a measured increase in reference temperature exceeds the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Revision 2, or if a measured decrease in Ch'arpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.1.2 in Regulatory Guide 1.99, Revision 2, the licensee is requested to report the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 16, 1991, and for the end of its current license.
Res onse 3.a The St. Lucie Unit 1 reactor vessel has not operated below 525'F.
as measured by the cold leg temperature for a significant period of time. The total cumulative time and reactor fluence accumulated at critical operations below 525'F..is conservatively estimated to be less than eight effective full-power hours or significantly less than 1 x 10" n/cm Since the resultant fluence is significantly below 1 x 10'/cm the effect of this low temperature irradiation on the reference g temperature and upper shelf is negligible. Additionally, the St.
Lucie Unit 1 Surveillance Capsule Reports have both representative plate, weld material, and SRM material, all of which have exhibited less transition shift that predicted by Regulatory Guide 1.99,
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Revision 2++ as described in Response 3c. Therefore, there is no evidence indicating the adverse affect associated with low temperature irradiation for the St. Lucie Unit 1 reactor vessel.
Res onse 3.b Florida Power & Light Co. responded to the NRC Generic Letter 88-11 with L-88-498+ for both the St. Lucie Units 1 and 2 reactor vessels. That letter stated that the P/T limit curves were in the process of being revised and would use the latest Regulatory Guide 1.99, Revision 2 methods for calculation of adjusted reference temperature of the reactor vessel beltline materials. On December 5g 1989 Florida Power 6 Light Co. submitted Amendment No. 104"+ for St. Lucie Unit 1, revising the P/T limit curves and LTOP setpoint analyses. The calculation of adjusted reference temperature used the prediction methodology in Regulatory Guide 1.99, Revision 2, Position 1.1, because the St. Lucie Unit 1 surveillance capsule data does not. meet the credibility requirement of having the limiting weld material in its program. The surveillance data has shown that the Regulatory Guide 1.99, Revision 2 does conservatively over predict the shift in adjusted reference temperature for the limiting plate, girth weld and SRM material.
Thus, the Regulatory Guide prediction of adjusted reference temperature for the limiting -beltline material used in the P/T limit curve and LTOP analyses is conservative and bounding when actual surveillance data is considered. On June 12, 1990, the NRC issued a letter that said the review of the new P/T limit curves (Amendment 104) were acceptable, which completed the review of the 88-11 issue for St. Lucie Unit 1.
Res onse 3.c Two surveillance capsules have been removed and tested from the St.
Lucie Unit 1 reactor vessel. The capsules tested to date cover an operation time span from start-up through part of 1990. The measured reference temperature for the St. Lucie Unit 1 surveillance capsule material limiting plate, intermediate to lower shell girth weld and SRM material exhibited an increase in adjusted reference temperature which was less than predicted by Regulatory Guide 1.99, Revision 2.
Measured reference temperature for the St. Lucie Unit 1 limiting weld material that has been irradiated in the Beaver Valley Unit 1 surveillance program has exhibited increases within the mean plus two standard deviations predicted by Regulatory Guide 1.99, Revision 2. Florida Power 6r Light Co. has plans to submit this data and its applicability to the St. Lucie'nit 1 surveillance program pending review of Reference 8.
The measured drop in upper shelf energy (USE) for the St. Lucie Unit 1 limiting base material from the first surveillance capsule exceeded the Regulatory Guide 1.99, Revision 2 value by 34 and 24 for the transverse and longitudinal orientation respectively. Test results from second capsule with higher fluence indicated the
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limiting plate longitudinal material had not exceeded the Regulatory Guide prediction. The measured drop in USE for the intermediate to lower shell girth weld material has not exceeded the Regulatory Guide prediction. The measured drop in USE for the St. Lucie Unit 1 limiting weld that has been irradiated in the Beaver Valley Unit 1 surveillance program did not exceed the prediction of the Regulatory Guide for all three capsules tested to date. The measured drop in USE for the SRM plate material from the second surveillance capsule exceeded the Regulatory Guide prediction by 10P. The first capsule did not have SRM material to confirm this as a "credible" trend.
As noted in the response to Response 2.a, the end of license (EOL) upper shelf energy for the limiting plate is not predicted to reach 50 foot-pounds at EOL. If an additional 34 reduction were applied to the Regulatory Guide 1.99, Revision 2 reduction in upper shelf to correspond to the 34 that the measured USE exceeded the Regulatory Guide prediction from the first capsule, there would still be adequate USE for the limiting plate and the EOL USE would not reach 50 foot-pounds as shown below (Case 1). This additional 3>o reduction is overly conservative, since results from the second capsule- with higher fluence showed USE drop to be less than the Regulatory Guide prediction.
Applying a 10% penalty (from the SRM test results) to the percent reduction for the St. Lucie Unit 1 limiting plate shows that there would still be an USE at EOL which is adequately above 50 foot-pounds (Case 2). This approach is considered overly conservative since the actual USE decrease for the vessel specific limiting plate and weld material from the second capsule, which yielded the SRM test result, were both less than predicted by Regulatory Guide 1.99, Revision 2.
No projections were made for the plate material at the December 16, 1991 date since the fluence is considerably less than that at the EOL, and projections at this time will obviously not drop below the 50 foot-pounds limit.
CASE MATERIAL INITIAL USE % Cu EQL 1/4 T RG 1.99 EOL USE Il-tb (1'raasvcrcc) Ha<<a<<c (a/cm') 't REDUCTION ft-lb Plate C-5935-2 (C-8-2) 103 0.15 '
'1 x 10 28+3 (31) 71. 1 Plate C-5935-2 (C-8-2) 103 0.15 2.01 x 10'8+10 (38) 63.9
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TABLE 1 ST LUCIE UNIT 1 REACTOR VESSEL BELTLINE PLATE MATERIAL DROP (6)
WEIGHT TEST RTmT LOCATION HEAT NO CODE S Cu(6) % Ni(6) p(6) g S(6) ('F.) (F')
Intermediate Shell A-4567-1 C-7-1 0. 11 0.64 0.004 0. 013 0 08 Plate Intermediate Shell B-9427-1 C-7-2 0.11 0.64 0.004 0.010 -30 pa Plate Intermediate Shell A-4567-2 C-7-3 0.11 0.58 0.004 0. 012 -30 pa Lower Shell C-5935-1 C-8-1 0. 15 0.56 0.006 0.010 -10 20~
Lower Shell- C-5935-2 C-8-2 0.15 0.57 0.006 0.010 10b 20b Lower Shell C-5935-3 C-8-3 0.12 0.58 0.004 0.010 0 08 NA Not Available a Determined using MTEB 5-2 b Surveillance Program Data(')
TABLE 2 ST. LUCIE UNIT 1 REACTOR VESSEL BELTLINE WELD MATERIAL DROP WEIGHT RTxoT WELD LOCATION HEAT NO FLUX TYPE FLUX LOT % Cu % Ni TEST (oF.) ('F.)
Intermediate Shell A8746/ Linde 124 3878/ 0.12(11) 0.20c 0. 018(11) 0. 017(12) NA -56a Long Seam (2-203 34B009 3688 Ag B, C)
Lower Shell 305424 Linde 3889 0 30(12) p 64(12) p 013(12) 0. 010(~2) NA -56a Long Seam (3-203 1092 A, Bg C Intermediate to Lower Shell Girth 90136 Linde 0091 3999 0 '3 O. 1,1b 0.013b O.O12b -6ob -6ob Seam (9-203)
NA Not Available a Generic data for CE submerged are welds using Linde 0091, 1092 and 124 Flux per 10 CFR 50.61 and Ref. 9 b Surveillance Program Data(')
c Estimated Ni content (low nickel type wire)(")
I REFERENCE LIST "Florida Power & Light Co., St. Lucie Unit 1, Updated Final Safety Analysis Report", Chapter 5.0 "Safety Evaluation of the St. Lucie Plant Unit No. 1", U.S.
Atomic -Energy Commission, Directorate of Licensing, November 8, 1974, Section 5.3 (3) "Florida Power & Light Co. St. Lucie Unit 1 Evaluation of Base Line Specimens", Combustion Engineering, Inc., October 1984, TR-F-MCM-005 "Analysis of Capsule W from the Duquesne Light Co. Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp., November 1988, WCAP-12005 "Florida Power & Light Co. St. Lucie Unit 1 Post Irradiation Evaluation of Reactor Vessel Surveillance Capsule W-97",
Combustion Engineering, Inc., December 1983, TR-F-MCM-004 (6) "Summary Report on Manufacture of Test Specimens and Assembly of Capsules for Irradiation Surveillance of Hutchinson Island Plant Unit 1 Reactor Vessel Materials", Combustion Engineering, Inc., July 1972, CENPD-39 "Analysis of the Capsule 104'rom the Florida Power & Light Co. St. Lucie Unit 1 Reactor Vessel Radiation Surveillance Program", Westinghouse Electric Corp., November 1990, WCAP-12751 (8) "Application of Reactor Vessel Surveillance Data for Embrittlement Management", ABB-Combustion Engineering, Inc.,
November 1991, CEN-405-P, Rev. 1-P FPL Letter, L-88-498, St. Lucie Units 1 and 2 Response to Generic Letter 88-11, W. F. Conway to NRC, November 23, 1988 (10) FPL Letter, L-89-408, St. Lucie Unit 1 Proposed License Amendment, P/T Limits and LTOP Analysis, J. H. Goldberg to NRC, December 5, 1989 (11) "Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCA's with Loss of Feedwater for the Combustion Engineering NSS", Combustion Engineering Owners Group, December 1981, CEN-189 and CEN-189 Appendix F (12) FPL Letter, L-77-308, St. Lucie Unit 1 Reactor Vessel Material Information, R. E. Uhrig to D. K. Davis, NRC, September 30, 1977
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