L-97-136, Forwards Response to NRC 970313 RAI on Pressurized Thermal Shock Submittals for Plant Rv Beltline Matls Per 10CFR50.61
| ML17229A347 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/16/1997 |
| From: | Stall J FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| L-97-136, TAC-M95484, TAC-M95485, NUDOCS 9705220305 | |
| Download: ML17229A347 (14) | |
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CATEGORY 1 REGULATORY INFORMATION DISTRIBUTION STEM (RIDS)
ACCESSI(,N NBR:9705220305 DOC.DATE: 97/05/16 NOTARIZED:
NO FACIAL:50-335 St. Lucie Plant, Unit 1, Florida Power
& Light Co.
50-389 St. Lucie Plant, Unit 2, Florida Power 6 Light Co.
AUTH.NAME AUTHOR AFFILIATION S'PALL,J.A.
F1o rida Power-,,6 Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
DOCKET 05000335 05000389
SUBJECT:
Forwards response to NRC 970313 RAI on pressurized thermal shock submittals for plant RV beltline matls per 10CFR50.61.
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. 13
Florida Power &Light Company, 6501 South Ocean Drive, Jensen Beach, FL34957 May 16, 1997 L-97-136 10 CFR 50.4 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 RE:
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 NRC TAC Nos. M95484 and M95485 Request for Additional Information - Response 10 CFR 50.61 - Pressurized Thermal Shock Evaluation The attached provides the response to the NRC request for additional information (RAI), dated March 13, 1997, on the 10 CFR 50.61 pressurized thermal shock (PTS) submittals for the St.
Lucie Unit 1 and 2 reactor vessel (RV) beltline materials.
The attached response supports the FPL conclusion that the 10 CFR 50.61 values of RT~ for each of the reactor vessel beltline material remain below the regulatory limits through the end of license (EOL). A change to a generic initialRT~ value (material property) for a nonlimiting RV weld for St. Lucie Unit 2 is also provided at the NRC's request.
The 10 CFR 50.61(b)(1) PTS evaluations were submitted by Florida Power and Light Company (FPL) letter, L-96-112, on May 14, 1996, and supplemented by FPL letters, L-96-233 and L 10, on September 23, 1996, and January 14, 1997, respectively The supplements were submittted in response the NRC requests on August 27, 1996, and October 15, 1996.
The evaluations determined that the projected reference temperature (RT~) at EOL for the reactor vessel beltline materials of each reactor vessel is acceptable.
FPL originally requested NRC approval of these evaluations by April 1, 1997, to support design of the fuel for St. Lucie Unit 1 Cycle 15. However, due to outage schedule changes for Cycle 15, FPL was able to revise the requested approval date to April 1, 1998.
This date supports the fuel design for St. Lucie Unit 1 Cycle 16.
This letter does not contain any new regulatory commitments.
Please contact us should you need any additional information to support your assessment.
Very truly yours, J. A. Stall Vice President St. Lucie Plant JAS/GRM 9705220305 9705ib
- PDR, ADOCK 05000335 P
PDR cc:
Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant an FPL Group company lllllllllllllllllllllllllllllllllillllll
~St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-136 Attachment Page 1
Response to Nrc Supplemental Request for Additional Information St. Lucie Units 1 and 2 Pressurized Thermal Shock Evaluation The Nuclear Regulatory Commission staff requested additional information from Florida Power and Light Company on the pressurized thermal shock evaluation'.
The questions and the responses are as stated below:
Question 1:
The licensee's response to Question 2 of the staff's Request for Additional Information (RAI) provided the basis for determining initial RT~~ (RT~Y~) values for several weld wire heats.
The licensee's responses with respect to three of the five heats were acceptable, However, the following information is necessary to complete our evaluation of the remaining two heats:
a)
The limitingweld in the St. Lucie 1 reactor vessel beltline is fabricated from weld wire heat 305424.
This heat of weld wire was also used to fabricate the Beaver Valley 1 surveillance weld and welds in the LaSalle 1 reactor vessel beltline.
The licensee's response 2 to the RAI indicated that the LaSalle and St. Lucie 1 Charpy data were not used to assess the RT~~<z for the St. Lucie 1 vessel.
Explain the effect on the value of RT~Y~ that would result from including the Charpy data from St. Lucie 1 and LaSalle 1.
Verify whether the RT~~<z value for heat 305424 remains drop weight controlled (i.e., does the lower bound Charpy curve become controlling).
b)
With regard to heat 83642, St. Lucie 2 reported an RT~~~ value of -80 F; and Beaver Valley 2 reported a value of-30'F.
The response stated that an RT>>Y~ of -80'F would be used for the St. Lucie 2 weld.
Provide the basis for selecting the non-conservative value of-80'F. Ifjustification cannot be provided, use a generic value in which plus or minus 2 sigma would bound the St. Lucie 2 and the Beaver Valley 2 data points.
Response to Question 1a:
The RT~~ was determined from drop weight test results of -60'F NDTI'nd compared to a full Charpy curve from data produced as part of the initial property testing for The Beaver Valley Unit 1 (BV-1) surveillance program'.
This was the only measured drop weight and complete Charpy data available for weld wire heat 305424 with Linde 1092 flux. This weld heat and flux combination was also used to fabricate the St. Lucie Unit 1 and LaSalle Unit 1 vessel beltline welds.
Three Charpy tests were performed at +10'F for the weld heat, 305424 with the same Linde 1092 flux, used to fabricate the St. Lucie Unit 1 and LaSalle Unit 1 vessel beltline welds.
These three Charpy test data4 of 82, 87, and 92 ft-lbs were reported (No lateral expansion or shear data was recorded) for both the St. Lucie Unit 1 and LaSalle Unit 1 vessel beltline welds.
The available industry Charpy Data'4 for Heat 305424 is presented in the table below:
~St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389
,L-97-136 Attachment Page 2 Charpy Impact Data for Heat 305424 with Linde 1092 lot 3889 Flux Specimen ¹
/Source 1 / BV-1 2/BV-1 3 / BV-1 4/BV-1 5 / BV-1 6 / BV-1 7/BV-1 8/BV-1 9/BV-1 10 / BV-1 11 / BV-1 12 / BV-1 13 / BV-1 14 / BV-1 15 / BV-1 16 / BV-1 17 / BV-1 18 / BV-1 1/ 10'F Data 2/ 10'F Data 3/ 10'F Data Test Temp.
( F)
-150'F
-150'F
-150'F 40'F
-60'F
-60'F
-25'F
-25'F
-25'F OOF O'
OOF 100'F 100'F 100'F 210'F 210'F 210'F
+10'F
+10'F
+10'F Impact Energy (ft-lb) 2.5 37 27 26 88 77 75 80 66.5 88 108.5 117.5 110 103.5 122 82 87 92 Lat. Exp.
(mils) 28 22 22 68 58 59 57 47 60 78 81 88.5 84 82 93 NR NR NR
% Shear 50 40 25 35 35 30 85 70 70 75 50 75 95 99 NR NR Figure 1 shows a plot of the 18 BV-1 baseline surveillance Charpy data points using a hyperbolic tangent curve fit. Figure 2 shows a plot of the 18 BV-1 baseline surveillance Charpy data points with the 3 Charpy tests at + 10'F using a hyperbolic tangent curve fit. Comparison of these two graphs in Figures 1 and 2 show that the three +10'F Charpy data points; have little effect on the baseline surveillance data; are above a lower bound Charpy curve that would be drawn through 66.5 ft-lb data point at O'; and that at NDTl'60'F) plus 60'F, or O', the drop weight temperature is still controlling for determination of RT~T<< by meeting the not less than 50 ft-lb requirement of NB-2331.
~$t. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L 97-136 Attachment Page 3 Response to Question 1b:
The RT~ of-80'F reported for heat 83642/Linde 0091/Lot 3536 as stated previously is a St.
Lucie 2 (SL-2) plant specific value that meets the requirements of NB-2331. The Beaver Valley 2 (BV-2) RT~T+ of -30'F and generic AT+
value of -56'F have been considered by calculating the effect these values of RTNDT(to have on EOL RTFI,. The EOL R@8 results of using these different RT~T~ values as compared to the St. Lucie Unit 2 limiting material are presented in the table below:
MATERIAL LOCATIONdc (CODE No.)
Data Source Int. Shell Axial Wclds (101-124 A,B,C)
SI 2S ficRT Int. Shell Axial Wclds (101-124 A>B>C)
Generic RT Int. Shell AxialWclds (101-124 A,B,C)
BV-2 RT Int. Shell Plate(M-60S-2) SI 2 Limiting Material HEATS
/LOT//
83642
/3S36 83642
/3636 83642
/3536 B3416-2 Chem.
Factor (cF) 30.7 30.7 30.7 91.S Imt.
-80oP
-S6'P
-30'F
+10'P Margin S6'P 66OF S60F 34OF EOL Hucnce n/cm'.76E19 2.76E19 2.76E19 2.76E19 hRTrss 39'P 39OF 390F 116'P EOL RTrss 1SoP 49oP 6S'F 1604P PTS Umit 270oF 270OF 270OF 270oP The above table shows that even using the most limitingBeaver Valley Unit 2 RT~To/, value of -30'F to calculate EOL RT~ values for the St. Lucie 2 intermediate shell axial welds, this material is over 200'F below the 10CFR50.61 PTS screening limitand 95'F below the most limitingSt. Lucie Unit 2 plate material.
However, since there is a large difference in these two RT~T(z test values and to ease the NRC review, the conservative generic value of -56'F with the larger margin term willbe used for calculations of EOL RT~8. This change willbe noted in Table 1 and Table 2 from Attachment B pages B-4 and B-S, respectively of the original PTS submittal'.
These changes willnot result in making this weld material limiting for EOL RTF13 or pressure temperature limitcurves.
Question 2:
The licensee's response to Question 3 of the RAI stated that "the fluence at the St. Lucie 1 limitingweld...has been updated."
The fluence value was 1.20 E19n/cm'n the original submittal.
Table 3 of the response to the RAI shows a value of 1.06 E19n/cm'.
Provide supporting documentation that justifies the decrease in the fluence value for St. Lucie 1. This includes an explanation of the analysis that was used to determine the revised fluence.
~St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L 97-136 Attachment Page 4 Response to Question 2:
The fluence value of 1.20 E19n/cm't 11.27 effective full power years (EFPY) for St. Lucie Unit 1 was the result of a typographical error. The data was mistakenly taken from data for the St. Lucie Unit 2 which utilized a vessel maximum fluence for all calculations since the Unit 2 vessel is less radiation sensitive.
This typographical error did not result in a change to end of license (EOL) RT~ calculations because the correct St. Lucie Unit 1 EOL fluence data was used for the RT~ projections.
This incorrect fluence data (flux/fluence rate) was only used as a comparison with the St. Lucie Unit 1 and Beaver Valley Unit 1 surveillance capsules.
The correct fluence for the St. Lucie Unit 1 vessel limiting weld is 1.06 E19n/cm's was reported in the RAI Response'.
The methodology to determine the St. Lucie reactor vessel neutron flux (E>1.0MeV) is based on computations of a computer codes package.
The codes package consists of the transport code, DOT4.3, and the core physics code, SIMULATE3. A linkage code, SORREL, has been used to connect DOT4.3 and SIMULATE3. A synthesis 3-D flux technique is used to provide the flux at the vessel wall. The basic nuclear data is based on the BUGLE-80 cross section library. Use of the codes package and the synthesis 3-D flux technique willgenerate the cycle specific flux data in question.
A product of the computed cycle specific flux and the plant cycle operation time provides the plant specific and cycle specific vessel fluence (E>1.0MeV) in question.
The fluence methodology has been extensively benchmarked against the measured data taken from several in-vessel and ex-vessel neutron dosimetry measurements conducted at Turkey Point Units 3 and 4 and St. Lucie Units 1 and 2 since 1984.
Question 3:
Where applicable, update the RT~ Tables as described in the response to the RAI.
Specifically, Tables 1 and 2 from Attachment B pages B-4 and B-5, respectively and Table 3 of Attachment A, page A-13.
Response to Question 3:
Changes have been made to Tables 1 and 2 from Attachment B pages B-4 and B-5, respectively of the original PTS submittal's indicated in the response to question 1b above.
The bolded changes also include those changes that were provided in a response to a previous NRC request for additional information'.
There are no additional changes to Table 3 of Attachment A, page A-13 other than those provided in reference 2.
~St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389
.L-97-136 Attachment Page 5 TABLE1:
St. Lucie Unit 2 Reactor Vessel Beltline Material Initial Properties MATERIAL LOCATIONtk
~ (CODE NO.)
Lower Shell Plate(M-4116-1 Lower Shell Plate(M-4116-2
'owerShell Plate(M-4116-3 Inter. Shell Plate (M-605-1 Inter. Shell Plate (M-605-2 Inter. Shell Plate(M-605-3 Inter. Shell Axial Welds (101-124 A,B,C Inter. Shell Axial Welds (101-124C)
Lower Shell Axial 101-142 A,B,C Inter. to Lower Shell Girth Weld (101-171)
Inter. to Lower Shell Girth Weld (101-171)
Surveillance Weld represents Weld (101-171)
HEATNO B-8307-2 A-3131-1 A-3131-2 A-8490-2 B-3416-2 A-8490-1 83642 83637 83637 3P7317 83637 83637 FLUX TYPE/LO T
NA NA NA NA NA NA Linde 0091
/3536 Linde 0091
/1122 Linde 0091/1122 Linde 124/0951 Llnde 124/0951 Linde 124/0951
% Cu 0.06 0.07 0.07 0.11 0.13 0.11 0.04 0.04 0.05 0.07 0.07 0.07
%Ni 0.57 0.6 0.6 0.61 0.62 0.61 0.07 0.07 0.1 0.08 0.08 0.08 INITIAL RTNDT (6
+20
+20
+20
+30
+10 5644 (Generic)
-50
-50
-80
-70
- This most limitingvalue applicable for weld 101-171.
- This conservative generic value used instead of measured data.
Changes to original submittal'able show in bold
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-136 Attachment Page 6 Table 2: St. Lucie Unit 2 Reactor Vessel Beltline Material EOL RT Values.
MATERIAL LOCATION& (CODE NO.)
Lower Shell Phte(M-4116-1)
Lower Shell Plate(M-4116-2)
Lower SheH Plate(M-4116-3 Inter. Shell Plate (M-605-1)
Inter. Shell Plate (M-605-2 Inter. SheH Plate (M-605-3 Inter. Shell AxialWelds (101-124 A,B,C)
Inter. SheH AxialWeld (101-124C)
Lower Shell AxialWelds (101-142 A,B,C Inter. to Lower Shell Girth Weld(101-171)
Inter. to Lower Shell Girth Weld(101-171)
Inter. to Lower Shell Surveillance weld HEAT NO B-8307-2 A-3131-1 A-3131-2 A-8490-2 B-3416-2 A-8490-1 83642 Linde 0091 Lot 3536 83637 Linde 0091 Lot 1122 83637 Linde 0091 Lot 1122 83637 Linde 124 Lot 0951 3P7317 Linde 124 Lot 0951 83637 Linde 124 Lot 0951
%Cu 0.06 0.07 0.07 0.11 0.13 0.11 0.04 0.04 0.05 0.07 0.07 0.07 0.57 0.60 0.60 0.61 0.62'.61 0.07 0.07 0.10 0.08 0.08 0.08 Chemistry Factor (CP) 37.0 44.0 44.0 74.15 91.5 74.15 30.7 30.7 37.5 41.2 41.2 41.2 Initial RTNDT
+20'F
+20'F
+20'F
+30'F
+10'F OOF
-56'F
-50'F
-504F
-70'F 40'F
-504F*>>
Margin 34'F 34'F 34'F 34'F 34OF 34'F 66'F 56'F 56'F 56'F 56'F 564F EOL Fluence n/cm'.76E19 2.76E19 2.76E19 2.76E19 2.76E19 2.76E19 2.76E19 2.76E19 2.76E19 2.76E19 2.76E19 2.76E19 ARTrrs 474P 564P 56'F 940F 116'F 94'F 39'F 39'F 480F 52'F 520F 52'F EOL RTrrs 101'F 110 P 110'F 158'F 160OF 128'F 49OF 45'F 54'F 38'F 28'F 58'F FfS Limit 270'F 270'P 270'P 270'F 270'F 270OF 270'P 270'F 270'F 300'F 300'F 3004F
- Limitingproperty for the inter. to lower shell girth weld 101-171.
Changes to original submittal'able show in bold
St. Lucie Units 1 and 2
~Docket Nos. 50-335 and 50-389 i 97-136 Attachment Page 7 Linde1092 Heat 305424/ Lot 3889 baseline CYGRhPH 4J Hyperbolic Tangent Curve Printed at 14J260 on f0-31-1997 Page 1
Coefftcienla of Curve 1
B -"5I09 Pquation tv. CYN = A + B '
tanh((T - 70)/C) l Upper Shelf Energy.'06 Temp. at 30 ft-lie W8 Temp. at 50 ft-lh ~
lnNer Shelf Energy: 221 Materiah lfKLD HeaL Number. XM4/lot3M9 Orientation:
Capsule init Total Huenca 0 150 100 Temperature
-150
-1M
-150
-25 Mfcrcntial
-89 13
-J9 HN BJ2
-3N 23 153 43
~ nate centlnwd on next page
~'200
-100 0
100 I 300 400 500 Temperature in Degrees F
Ihta Sells) Plotted Plant:
BV1 Cap: init Materiat lfElLI Ori Heat f: X5424/lot3689 Charpy V-Notch Data input CYN Energy Compu4d CYN Energy 2
269 4
216 25 269 26 3087 37 XN 27 X87 75 7289 60 7289 77 7289 Figure 1: Plot ofWeld Heat 305424 with Linde 1092 lot 3889 Baseline Charpy Data from the Beaver Valley Unit 1 Surveillance Program.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-136 Attachment Page 8 Linde1092 Heat 305424/ Lot 3889 baseline Page 2 IIEt0 Heat Number. 30QH/lot3%9 Orientation:
Capsule init Total fluence 0 Char py V-Notch Data (Continued)
Temperature 0
0 0
100 l(8 IOO 210 210 210 Input CVN Energy 00 00 K5 1175 100 1065 122 110 1035 Camputat CYN KQ3 9333 9333 10589 NN lms 100 IOO loi Energy OifferenM
-1333
-2M3 IIS 26 1599 399
-25 SUII of RESIDUhls =
0 Figure 1 Continued: Plot of Weld Heat 305424 with Linde 1092 lot 3889 Baseline Charpy Data from the Beaver Valley Unit 1 Surveillance Program.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-136 Attachment Page 9 Linde Ht 305424/Lot3889 baseline w/ IOF CYCRAPH 41 Hyperbolic Tangent Curve Printed at IQM1 on 03-31-1997 Page I Coefficients of Curve 1
h = 5328 B = 5123 Eciuatlon h CYH = h 4 B '
tanh((T - TD)/C) l Upper Shelf Energy: 1MSI Temp. at 30 ft-Ibs HL4 Temp at 50.ft-lb' Lower Shelf Energy: 204 Lsteriab IIELD Heat Number. 305124/Iot3889 Orientation:
Capsule init Total Fluence 0 Ing 150 l4 100 Temperature
-IM
-150 "L50
-25 "25
-25
-200
-100 0
100 200 400 Temperature in Degrees F
Data Sells) Plot&
PLmL BY1 C'ap. init HateriaL IfRU)
Orl:
Heat f. 305424/Iot3IS Char py V-Ncotch Data Input CYN Fzergy Computel CYN Energy 2
2,71 2.71 25 271 26 3I37 37 3L37 27 3IB7 75 70I8 88 70I8 77 7018 Differential
-.71 128 "21 M2 437 4BI 17BI 6BI
~" Doto conunucd on next paac ~"
Figure 2: Plot ofWeld Heat 305424 with Linde 1092 lot 3889 Baseline Charpy Data from the Beaver Valley Unit I Surveillance Program with 10nF Data.
St. Lucie Units 1 and 2
'Docket Nos. 50-335 and 50-389 I 97-136 Attachment Page 10 Linde Ht 305424/Lot3889 baseline w/ 10F Page 2 WELD Heat Humber. 305424/lotm9 Orientation:
Capsule init Total Fluence 0 Charpy V-Notch Data (Continued)
Temperalure 0
0 0
10 10 10 I00 I00 IOO 210 210 210 Input CYN Energy 88 80 K5 87 82 92 I00 10&5 1175 1035 122 110 Computed CYH 9031 9131 9031 95N 95N 95N I%34 l0434 IOI34 I045l 10451 IM51 Energy Differential MI
-1031
-2381
-13N
-3N 415 13.15
-101 1748 5.48 SUM of RESIUhLS =
0 Figure 2 Continued: Plot of Weld Heat 305424 with Linde 1092 lot 3889 Baseline Charpy Data from the Beaver'alley Unit 1 Surveillance Program with 10'F Data.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-136 Attachment Page 11 1.
FPL Letter, L-96-112, "St. Lucie Unit 1 and 2 Docket No. 50-335 and 50-389, 10 CFR 50.61 Evaluation of Pressurizer Thermal Shock of Reactor Vessel Beltline Materials",
W.H. Bohlke to NRC, May 14, 1996.
3.
"Duquesne Light Co. Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp.,
October 1974, WCAP-8457.
4.
"AtypicalWeld Material In Reactor Pressure Vessel Welds" CE Response to I&EBulletin 78-12, June 8, 1979,Section VIII,Page 22.
(SW 2.
FPL Letter, L-97-10, "St. Lucie Unit 1 and 2 Docket No. 50-335 and 50-389, Request for Additional Information (RAI) - Response, 10 CFR 50.61 - Pressurized Thermal Shock Evaluation", J. A. Stall to NRC, January 14, 1997.
1 t
V l