ML17229A201
| ML17229A201 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 01/14/1997 |
| From: | Stall J FLORIDA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| L-97-10, TAC-M95484, TAC-M95485, NUDOCS 9701210450 | |
| Download: ML17229A201 (20) | |
Text
CATEGORY 1 REGULAT INFORMATION 'DISTRIBUTION STEM (RIDS) r'.
ACCESSION NBR:9701210450 'OC.DATE: 97/01/14 NOTARIZED: NO
'ACIL:50-335 St. Lucie Plant, Unit'1, Florida Power a Light Co.
50-389 St. Lucie Plant, Unit 2, Florida Power
& Light Co.
AUTH.NAME AUTHOR AFFILIATION STALL,J.A.
Florida Power
& Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards response to RA'I re pressurized thermal shock evaluations.
DISTRIBUTION CODE:
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TITLE: OR Submittal:
Thermal Shock to Reactor Vessel DOCKET g'5000335 05000389 NOTES:
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E NOTE TO ALL "RIDS" RECIPIENTS:
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Florida Power 8 Light Company, P.O. Box 128, Fort Pierce, FL 34954-0128 January 14, 1997 L-97-10 10 CFR 50.4 10 CFR 50.61 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 (RAI) - Response 10 CFR 50.61 - Pressurized Thermal Shock Evaluation By letter dated October 15, 1996, the NRC staff requested additional information needed to complete their assessment of the pressurized thermal shock (PTS) evaluations for St. Lucie Units 1 and 2. The response to your RAI is attached.
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 letter, L-96-233, on September 23, 1996, The evaluations determined that the projected reference temperature (RT~)
at the end of license (EOL) for the reactor vessel beltline materials of each reactor vessel.
The EOL RT~ values were compared against the regulatory limits and determined to be acceptably below those limits. FPL originally requested NRC approval of the PTS evaluations by April 1, 1997, to support design of the fuel for St. Lucie Unit 1 Cycle 15.
Due to outage schedule changes for Cycle 15, FPL now requests approval by April 1, 1998, to support 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 9] Oo 49,
/PAL cc:
Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant 970i2i0450 970ii4 PDR ADQCK 05000335 P
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page
1 Background
The NRC request for additional information generated from the Florida Power and Light Company pressurized thermal shock evaluations's restated below:
The staff requires additional information in order to complete its review of the Florida Power and Light Company pressurized thermal shock (PTS) evaluations.
The submittal consisted of three parts:
"St. Lucie Unit 1 Reactor Vessel Beltline Materials End of License Assessment of RT~"
"St. Lucie Unit2 Reactor Vessel Beltline Materials End of License Assessment of RT~"
Proprietary and non-proprietary copies of the Combustion Engineering Owner's Group (CEOG) report CEN-405-P, Revision 3, "Application of Reactor Vessel Surveillance Data for Embrittlement Management" Below are the four NRC questions with the FPL response provided after each question:
NRC Request 1:
Provide the actual cold leg temperatures for each cycle for both St. Lucie Unit 1 (SL-1) and Beaver Valley Unit 1 (BV-1) in support of the integrated surveillance approach.
Page A-7 of the submittal discusses "nominal inlet temperatures" as listed in the updated final safety analysis report (UFSAR) and "assumed cold leg temperatures."
The actual cold leg temperatures for each cycle are needed for both units in order for the staff to determine the applicability of the BV-1 surveillance data to the SL-1 reactor vessel.
FPL Response 1:
The Beaver Valley Unit 1 (BV-1) surveillance specimens from capsules V, U and W were removed from the BV-1 reactor vessel at the end of cycle 1, cycle 4, and cycle 6 respectively.
In order to compare the surveillance specimen temperature environment, the actual cold leg temperature data from the BV-1 control room logs were reviewed.
The logs from cycle 4 through cycle 7 were made available by Duquesne Light Co. to Florida Power and Light. The logs were reviewed with minimum, maximum and bulk (majority of the data) temperature entries with greater than 95% power recorded for each cycle.
The minimum and maximum temperatures for each cycle were averaged to further reduce the data for comparison.
Although cycle 7 was after the BV-1 surveillance capsules were removed, it was evaluated to show uniformity with the T-cold data.
The results are provided in the following table.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 2 Beaver Valley Unit 1 Cold Leg (Inlet) Temperature Data Cycle Bulk of data centered Minimum Cold Lcg Maximum Cold Leg Average of Min.&,
Avcragc 4-7 544'F 544oP 5450 F 545 F 541.5oF 541.8'F 543.5'F 542.8'F 546.0'F 546.2'F 547.8'F 547.4'F 543.8'F 544.04P 545.74F 545.1'F 544.7'P The Beaver Valley cycle 1 through cycle 3 data were not available.
However, Beaver Valley plant personnel indicated that the BV-1 cold leg temperature had remained essentially unchanged from startup through the cycle 7, and that the data from cycle 4 through cycle 7 above should be representative for cycle 1 through cycle 3'. Therefore, based on the actual temperature data from cycle 4 through cycle 7, the BV-1 cold leg inlet temperatures representative of cycle 1 through cycle 7 were assumed to be 544.7'F.
The St, Lucie Unit 1 actual cold leg temperature data from the control room logs were reviewed from the end of cycle 4 through cycle 13. Since the temperature data were uniform and there was a large amount of data, full power steady state temperatures were representative of the beginning and end of each cycle.
The actual cycle specific data are presented in the following table.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 3 St. Lucie Unit 1 Cold Leg (Inlet) Temperature Data Date Cycle A Cold Lcg 0
B Cold Leg 0
Avcragc Cold Lcg Temp.
0 a7/30/81 OC
- 4/1/82 OC 9/12/82 OC 2/10/83 OC 6/14/84 OC 9/14/85 OC 1/10/86 OC 1/20/87 OC 5/14/87 OC 7/8/88 OC 9/12/88 OC 1/9/90 OC 5/17/90 OC 10/17/91 OC 1/7/92 OC 3/27/93 OC 7/4/93 OC 10/24/94 OC 12/20/94 OC 4/10/96 OC Average - 2700 Mwt thermal
- Average - 2560 Mwt thermal 10 10 12 12 13 13 5-13 4, start of 5
541.4 541.4 548.3 548.8 548.5 548.8 549.4 548.9 549.3 549.0 549.0 549.5 549.2 549.2 548.9 549.1 549.0 549.2 549.2 548.3 541.7 541.3 547.4 548.3 548.4 548.6 548.3 548.7 548.2 548.4 548.5 547.9 548.5 548.4 548.7 548.3 548.7 548.8 548.8 548.1 541.6 541.4 547.9 548.6 548.5 548.7 548.9 548.8 548.8 548.7 548.8 548.7 548.8 548.8 548.8 548.7 548.8 549.0 549.0 548.2 548.7 541.5
- Prior to up rating to 2700 MWt on 4-10-82 BOC-Near the Beginning of the cycle EOC-Near the End of the cycle The temperature data for St. Lucie Unit 1 cycle 1 through cycle 3 and the early part of cycle 4 were not readily available.
From cycle 1 through cycle 4, St. Lucie Unit 1 operated with a 2560 MWt thermal core.
Plant personnel indicated that vessel inlet temperature for the end of cycle 4 and the beginning of cycle 5 (prior to the power up rate on April 10, 1982) should be representative of cycle 1 through cycle 4 vessel inlet temperature.
Therefore, the average reactor
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389
'-97-10 Attachment Page 4 vessel inlet temperature for St. Lucie Unit 1 for cycle 1 through cycle 4 was 541.5'F and for cycle 5 through cycle 13 was 548.7'F.
The time weighted average inlet temperature is determined as follows.
Using a cold leg inlet temperature of 541.5'F for 32,800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> (Cycle 1-4) and 548.7'F for the remaining 85,600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> (Cycle 5-13) the time weighted average SL-1 nominal cold leg inlet temperature is 546.7'F which is almost identical to the 544.7'F BV-1 cold leg data.
Since the BV-1 surveillance capsules were irradiated at a similar, but slightly lower c'old leg temperature compared to the St. Lucie Unit 1 cold leg temperature, the BV-1 surveillance data does not require any temperature correction for use.
The inlet temperature comparison between the BV-1 surveillance capsule specimens and the SL-1 reactor vessel meets the credibility criteria of + 25'F irradiation temperature requirement stated in 10 CFR 50.61.
NRC Request 2:
Determine ifthere are other initial test data (Charpy, Drop Weight and RT>>~z, values) for heats 305424, 90136, 83642, 83637 and 3P7317.
Provide a listing of each initial reference temperature value, including the source of the data. Ifother data are available, justify the initial reference temperature values that appear in Table 4 (page A-14) and Table 2 (page B-5) of the submittal for the heats listed above.
FPL Response 2:
To determine ifthere are other initial test data for the above listed weld wire heats applicable to the St. Lucie Units 1 and 2, the Electric Power Research Institute (EPRI) RPVDATA(Version 1.3) database was reviewed.
The EPRI RPVDATA database contains an NRC Best Estimate Data screen of the results of the NRC RVID database from the Generic Letter 92-01 responses.
St. Lucie Unit 1 was designed and built to ASME Section III 1965 Edition, Winter of 1967 Addenda when drop weight testing of weld metal was not required.
At that time it was the practice for Combustion Engineering (CE), the nuclear steam supply system vendor, to perform weld material drop weight testing on the weld material in the reactor vessel surveillance program.
As a result, there is usually a generic value or only one initial RT>>T test value for each weld wire heat.
This vessel was built at a time when the weld wires were copper coated, and nickel was alloyed with some of the wires.
Therefore, the welds are the limiting material and the initial RT>>T values are critical in the determination of embrittlement predictions such as PTS.
St. Lucie Unit 2 was designed and built to ASME Section III 1971 Edition, Summer of 1972 Addenda, when both drop weight and Charpy testing was required for all pressure boundary materials, Therefore, there is measured initial RT>>T data for every heat of weld material and in some cases, multiple data exist.
The St. Lucie Unit 2 weld heat initial properties are not critical
4 y
t 4
"k iJ
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 5 for embrittlement predictions since these low copper, low nickel welds are not the limitingvessel material and their end of license shift in RT>>T is less than 60'F.
The St. Lucie Unit 2 vessel also has some welds that used multiple wire heats, in one case as a result of a repair.
In cases with multiple heats of wire, FPL conservatively chose the higher of the initial RT>>T values and assigned it to the weld seam.
When this weld seam was reported in a database, the higher "conservative" value was sometimes assigned to both heats separately.
This can add to the variability of the data outside of the normal variability of the measured test data.
Table 1 and Table 2 now list each heat separately with the specific initial reference test value
'eported.
The results for each heat are discussed below.
St. Lucie Unit 1 Weld Wire Heats:
Heat 305424/Linde 1092/Lot 3889: This limitingweld heat was used in the fabrication of the St. Lucie Unit 1 vessel lower shell axial welds, 3-203A, B & C. The database search indicated there are three plants that have this wire heat in combination with Linde 1092 flux, all with the same flux lot. The initialRT~ of-60'F is based on actual drop weight and Charpy results from the Beaver Valley Unit 1 surveillance program'.
The only other initial RT>>T values reported are a generic value of -56'F and a plant specific -50'F from LaSalle Unit 1.
The -50'F LaSalle value is not a measured value based on drop weight testing.
It is a calculated result based on
+10'F Charpy data as identified in the LaSalle Unit 1 response to Generic Letter 92-01'nd was submitted as a licensed value specific for LaSalle.
Therefore, the initial RT>>T value of -60'F is justified as the only measured value that meets the requirements of ASME Section IIINB 2331.
Heat 90136/Linde 0091/Lot 3999: This non limitingweld heat was used in the fabrication of the St. Lucie Unit 1 vessel intermediate to lower shell weld, 9-203, and the surveillance weld.
The database search indicated there are 4 plants that have this wire heat in combination with different lots of Linde 0091 flux. These plants have either reported plant specific values of -60'F or the generic value ifthey have no data.
The St. Lucie Unit 1 value of -60'F is justified as a measured value that meets the requirements of ASME Section III NB 2331 with all other measured value in agreement.
St. Lucie Unit 2 Weld Wire Heats:
Heat 83642/Linde 0091/Lot 3536: This non limitingweld heat was used in the 'fabrication of the St. Lucie Unit 2 vessel intermediate shell axial welds, 101-124 A, B, & C. Allthree seams used this wire/flux/lotcombination, however, the C seam was repaired with another wire heat identified below.
The database search indicated one other plant that has this wire heat in combination with Linde 0091 flux. St. Lucie Unit 2 reported an initial RT>>T of -80 F based on the CE weld material qualification report'.
Beaver Valley Unit 2 reported an initial RT>>~ of-30'F based on a plant specific surveillance weldment tests, Both of these test values are valid data that meet the requirements of ASME Section IIINB 2331 and are within the ranges of reported
4 ilf II
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 6 data that were used to determine the generic RT>>~ value of -56'F for Linde 0091 flux welds'.
This variation should be of little concern for this low copper, low nickel weld, since it is more than 250'F below the 10 CFR 50.61 screening limitand more than 140'F below the most limiting St. Lucie Unit 2 beltline material at the end of license.
Even substituting the generic value of-56'F or -30'F this material willnot become limiting. Since no guidance exists for treatment of multiple initial RT>>Y data, and this is a non limitingweld, the St. Lucie Unit 2 value of -80'F initial RT>>~ should be applicable for the intermediate axial shell welds 101-124 A &B since it meets the requirements of ASME Section IIINB 2331.
Heat 83637/Linde 0091/Lot 1122: This non limitingweld heat was used in the fabrication of the St. Lucie Unit 2 vessel lower shell axial welds 101-142A, B & C and the partial repair of the intermediate shell axial weld 101-124C only. The database search indicated one other plant that has this wire heat in combination with Linde 0091 flux. St. Lucie Unit 2 reported an initial RT>>Y of -50'F based on the CE weld material qualification report'.
This value of -50'F is in agreement with the value reported for the other plant for this heat in combination with Linde 0091 flux. An error was noted in the EPRI RPVDATAdatabase for the flux lot and initial RT>>~ value reported for the intermediate shell axial weld 101-124C.
The correct flux lot is 1122 and the correct initial RT>>~ of -50'F will be reported to the organization tasked with the EPRI RPVDATA database update.
Since the intermediate shell axial weld 101-124C was fabricated with a single arc, with both wire heat 83642/Linde 0091/lot 3536 with an initial RT>>~ of -80'F, and heat 83637/Linde 0091/lot 1122 with an initial RT>>Y of -50'F, the conservative higher value of -50'F is the more limitingfor this weld for embrittlement predictions.
The St. Lucie Unit 2 value of -50'F initial RT>>~ for the lower shell axial welds 101-142A, B,
&C and the intermediate shell axial weld 101-124C is justified as a measured initial RT>>~ value that meets the requirements of ASME Section III NB 2331, with all other measured values in agreement.
Heat 83637/Linde 124/Lot 0951: This non limitingweld heat was used in the partial fabrication of the St. Lucie Unit 2 vessel intermediate to lower shell girth weld, 101-171, and all of the surveillance weld. The database search indicated that this wire heat in combination with Linde 124 flux was not used at any other plant.
The initial RT>>y value reported by CE was -70'F.
The same heat and flux combination was used to fabricate the surveillance weld.
The initial RT>>~ determined from the surveillance baseline specimen weld was -50'F'nd meets the requirements of ASME Section IIINB 2331.
This higher value of -50'F is applicable for the intermediate to lower shell weld, 101-171 and is corrected in Table 1 and 2 from Attachment B pages B-4 and B-5'.
Heat 3P7317/Linde 124/0951:
This non limitingweld heat was used in the partial fabrication of the St. Lucie Unit 2 vessel intermediate to lower shell girth weld, 101-171.
The database search indicated that this wire heat in combination with Linde 124 flux was also used to fabricate one other plant.
Even though the flux lot is different, both plants have plant specific values of
-80'F reported for the initial RT>>Y.
However, since this heat was also used to fabricate the intermediate to lower shell weld with wire heat 83637/Linde 124/lot 0951 flux, the higher value
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 7 of-50'F is more limitingfor determining embrittlement predictions for the intermediate to lower shell girth weld, 101-171.
NRC Request 3:
Provide the neutron flux for the BV-1 surveillance capsule and the SL-1 vessel to clarify the conclusion that the neutron flux values are within an order of magnitude of each other.
FPL Response 3:
Table 3 of Attachment A, page A-13 of the subject PTS submittal'as been revised to indicate the last column as flux(E) 1MeV). The fluence at the St. Lucie Unit 1 limitingweld and vessel maximums has also been updated.
Comparison of the neutron flux data for the Beaver Valley Unit 1 surveillance capsules and the St. Lucie Unit 1 limitingweld shows that both are within the same order of magnitude.
NRC Request 4:
Provide details on the equivalence of the BV-1 neutron spectra to that of SL-1 to quantify the conclusion on page A-6 of the submittal that "the irradiation behavior of the BV-1 surveillance specimen [is] therefore comparable to the SL-1 limiting weld material."
FPL Response 4:
The equivalence of the BV-1 neutron spectra to the SL-1 neutron spectra can be determined by comparison of the neutron spectra analyses for the last surveillance capsule removed from both units.
The capsule analyses represent the most realistic/accurate comparison since they are measured data points.
Each capsule report identifies the flux and D.A. at the capsules and transports the measured capsule analysis to the reactor vessel base metal clad interface.
The BV-1 capsule spectra as compared to the SL-1 vessel wall at the peak and critical weld locations are provided in the following table:
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 8 Phut (rcf. noted)
BV-1 ~
BV-1 0 SL 1 lo SL-1 'O SL-1
'ocation Ccntcr of Ca sulew Vessel QO'ax flux Center of Ca sulc 104'csscl 15'critical weld Vessel 0'ax flux I Flux, n/cm'-sec (E) 1McV) 5.08E10 Av
. c clc 1-6 4.60E10 Av.c clc14 3.00E10 Av. c clc 1-9 1.97E10 Av. c clc 1-9 3.14E10 Av. c cle 1-9)
D,A./sec 8.38E-11 7.49E-11 4.52E-11 3.00E-11 4.71E-11 Damage Ratio, D.A./scc/Flux 1.65E-21 1.63E-21 1.51E-21 1.52E-21 1.50E-21 Regardless of the different nuclear data used in the discrete ordinate transport analyses and different reactor internals, it can be concluded that the irradiation behaviors of the BV-1 capsule, which contains the SL-1 critical weld material and the SL-1 vessel at the critical weld location (1.65E-21/1.52E-21) are similar within 9%.
Therefore, the BV-1 surveillance data are appropriately applicable to the SL-1 vessel data.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 9 TABLE 1 St. Lucie Unit 2 Reactor Vessel Beltline Material Initial Pro erties MATERIAL LOCATION& (CODE NO.
HEAT NO FLUX TYPE/LOT
%Cu
% Nl INITIAL RTNDT O
Lower Shell Plate 4116-I Lower Shell Plate 4116-2 Lower Shell Plate 4116-3 Inter. Shell Plate 405-I Inter. Shell Plate 405-2 Inter. Shell Plate 405-3 Inter. Shell Axial Wclds (101-124 A,B,C)
Inter. Shell Axial Wclds 101-124 C 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 reprcscnts Weld (101-171)
B-8307-2 A-3131-1 A-3131-2 A-8490-2 B-3416-2 A-8490-1 83642 83637 83637 3P7317 83637 NA NA NA NA NA Linde 0091/3536 Linde 0091/1122 Lindc 0091/1122 Linde 124/0951 Linde 124/0951 Lindc 124/0951 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 0.61 0.07 0.07 0.10 0.0$
0.08 0.08
+20
+20
+20
+30
+10
-80
-50
-50
-80
-70
-50>>
- This most limitingvalue applicable for weld 101-171.
Changes to the original submittal'able are shown in bold
~ I
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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L 97-10 Attachment Page 10 Table 2 St. Lucie Unit 2 Reactor Vessel Beltline Material EOL RT Values Material Location &(Code Lower Shell Plate (M%116-1 Heat No.
B-8307-2
%Cn 0.06 0.57 37.0
% Ni Chemistry Factor Initial RTNer
+20'F 34OF EOL Fluence 2
2.76E19 hRTFts 47OF EOL 101'F PI'S Lhnit 270'F Lower Shell Plate (MC116-2)
Lower Shell Plate (M%116-3)
A-3131-1 A-3131-2 0.07 0.60 0.07 0.60 44.0 44.0
+20'F
+20'F 34OF 34'F 2.76E19 2.76E19 56'F 56'F 110'F 110'F 270'F 270'F Inter. Shell Plate (M%05-I)
A-8490-2 0.11 0.61 74.15 30OF 34OF 2.76E19 94OF 158'F 270'F Inter. Shell Plate (M%05-2 B-3416-2 0.13 0.62 91.5
+10'F 34'F 2.76E19 116'F 160'F 270'F Inter. Shell Plate (M-605-3)
A-8490-1 0.11 0.61 74.15 00F 34'F 2.76E19 94'F 128 F 270 F Inter. Shell AxialWelds (101-124 A,B,C)
Inter. Shell AxialWeld (101-124 C) 83642 Lindc 0091 Lot 3536 83637 Linde 0091 Lot 1122 0.04 0.07 0.04 0.07 30.7 30.7
-80'F 50OF 56'F 56'F 2.76E19 2.76E19 39OF 39OF 15'F 45'F 270'F 2700 F Lower Shell AxialWelds (101-142 A,B,C) 83637 Linde 0091 Lot 1122 0.05 0.10 37.5 50OF 56'F 2.76E19 48'F 54OF 270'F Inter. to Lower Shell Girth Weld(101-171) 83637 Linde 124 Lot 0951 0.07 0.08 41.2
-70'F 56'F 2.76E19 52'F 38'F 300'F Inter. to Lower Shell Girth Weld(101-171)
Inter. to Lower Shell Surveilhnce weld 3P7317 Linde 124 Lot 0951 83637 Llnde 124 Lot 0951 0.07 0.07 0.08 0.08 41.2 41.2
+0OF
-50'F**
56'F 56'F 2.76E19 2.76E19 52OF 52OF 28'F 58'F 3004F 300'F
- Limitingproperty for the inter. to lower shell girth weld 101-171.
Changes to the original submittal'able are shown in bold.
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 11 Table 3 Beaver Valley Unit 1 Surveillance Capsule and St, Lucie Unit 1 Vessel Fast Neutron Fluence (E>
1 MeV) Com arison Plant Capsule Fluence (/),
2 EFPY EFPS
SI 1
SI I 97'044 Vessel maximum Vcsscl limiting welds 2.91 E18 6.54 E18 9.49 E18 5.50 E18 7.16 E18 1.72 E19 1.06 E19 1.16 3.59 5.89 4.67 9.515 14.84 14.84 3.66 E07 1.13 E08 1.86 E08 1.47 E08 3.00 E08 4.68 E08 4.68 E08 2.51 E18 1.82 E18 1.61 E18 1.18 E18 7.52 E17 1.16 E18 7.14 E17 7.95 EIO 5.78 EIO 5.11 EIO 3.73 E10 2.39 E10 3.68 E10 2.26 E10
/
EFPY= Effective full power years EFPS= Effective full power seconds Changes to the original submittal'able are shown in bold.
Endnotes FPL Letter, L-96-112, "10 CFR 50.61 Evaluation of Pressurizer Thermal Shock of Reactor Vessel Beltline Materials," W. H. Bohlke to NRC, May 14, 1996.
2.
Duquesne Light Letter, ND1MNE:7589, "Reactor Vessel Inlet Temperature for Beaver Valley Unit 1," L.R. Freeland to R. S. Boggs, FPL, November 7, 1996.
3.
"Duquesne Light Co. Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program,"
Westinghouse Electric Corp.,
October 1974, WCAP-8457.
4.
Commonwealth Edison Letter, ZNLD/1790, "Dresden..., QuadCities..., LaSalle County Units 1 and 2.... Response to Generic Letter (GL) 92-01, Reactor Vessel Structural Integrity, Revision 1," Marcia A. Jackson to Dr. Thomas E. Murley, Director, USNRC, July 1, 1992.
5.
"Atypical Weld Material In Reactor Pressure Vessel Welds," CE Response to I&E Bulletin 78-12, June 8, 1979,Section VIII,Page 61.
6.
"Evaluation of PTS Effects Due to Small Break LOCA's with Loss of Feedwater for the Combustion Engineering NSSS," Section 6.4, CE Owners Group, December 1981, CEN-189.
4 v
St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-10 Attachment Page 12 7.
"AtypicalWeld Material In Reactor Pressure Vessel Welds" CE Response to I&EBulletin 78-12, June 8, 1979,Section VIII,Page 69.
8.
9.
I "Analysis of Capsule W-83, Florida Power and Light Co. St. Lucie Unit No. 2",
Babcock &Wilcox, September
- 1985, BAW-1880.
"Analysis of Capsule W from Duquesne Light Co. Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp.,
November
- 1988, WCAP-12005, Table 6.1
~
10.
"Analysis of the Capsule at 104'rom the FPL St. Lucie Unit 1 Reactor Vessel Radiation Surveillance Program",
Westinghouse Electric Corp.,
November 1990, WCAP-12751.