ML20141K043
| ML20141K043 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 05/21/1997 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20141J982 | List: |
| References | |
| NUDOCS 9705280295 | |
| Download: ML20141K043 (23) | |
Text
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ATTACHMENT I BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT i
i 9705280295 970521-PDR ADOCK 05000454 P PDR,
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BRAIDWOOD - UNIT 2
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PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page j 1.0 Introduction 1 2.0 Operating Limits 1
2.1 RCS Pressure and Temperature (P/T) Limits 1 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints 2 2.3 LTOP Enable Temperature 2 3 2.4 Reactor Vessel Boltup Temperature 3
! 2.5 Reactor Vessel Minimum Pressurization Temperature 3 3.0 Reactor Vessel Material Surveillance Program 1 9 4.0 Supplemental Data Tables
. 11 5.0 References 18 Attachment WCAP-14824, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron and Braidwood," Revision 1, April 1997. '
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BRAIDWOOD - UNIT 2 i a !
l PRESSURE AND TEMPERATURE LIMITS REPORT !
List of Figures i
' I Figure I Page l 2cl Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 4 l
] 100 F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation :
Errors) l i
2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up 5 to 100 F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) j 2.3 Braidwood Unit 2 Maximum Allowable Nominal PORV Setpoints for the Low 7 Temperature Overpressure Protection (LTOP) System Applicable for the First 16 EFPY
)
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i BRAIDWOOD - UNIT 2 i PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1 Braidwood Unit 2 Heatup and Cooldown Data Points at 16 EFPY 6 (Without Margins for Instrumentation Errors) 2.2 Data Points for Braidwood Unit 2 Maximum Allowable PORV Setpoints 8 for the LTOP System Applicable for the First 16 EFPY 3.1 Braidwood Unit 2 Capsule Withdrawal Schedule 10 !
4.1 Braidwood Unit 2 Calculation of Chemistry Factors Using Surveillance 12 Capsule Data 4.2 Braidwood Unit 2 Reactor Vessel Material Properties 13 4.3 Summary of Braidwood Unit 2 Adjusted Reference Temperatures (ARTS) at 14 the 1/4T and 3/4T Locations for 16 EFPY 4.4 Braidwood Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 15 16 EFPY at the Limiting Reactor Vessel Material Weld Metal WF562 (Based on +
Surveillance Capsule Data) 4.5 RTns Values for Braidwood Unit 2 for 32 EFPY 16 l 4.6 RTns Values for Braidwood Unit 2 for 48 EFPY 17 l
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BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction Reference to Technical Specifications (TS) numbers are given in both the Braidwood Station current Technical Specifications (CTS) and Improved Technical Specifications (ITS). The CTS number is presented first, followed by the ITS number in brackets [ ].
This PTLR for Unit I has been prepared in accordance with the requirements of TS 6.9.1.11/[ITS-5.6.6]. Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications addressed in this report are listed below:
LCO 3.4.9.1 Pressure / Temperature Limits; and LCO 3.4.9.3 Overpressure Protection Systems.
[ITS-LCO 3.4.3 RCS Pressure and Temperature (Pff) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System].
l 2.0 Operating Limits The PTLR limits were developed using a methodology specified in the Technical Specifications.
The methodology listed in WCAP-14040-NP-A (Reference 1) was used with four exceptions:
- a. Use of ENDF/B-IV neutron transport cross-section library and ENDF/B-V dosimeter reaction Cross-seClions,
- b. Optional use of ASME Code Section XI, Appendix G, Article G-2000,1996 Addenda, I c. Use of ASME Code Case N-514, and
- d. Use of RELAP computer code for calculation of LTOP setpoints for Unit I replacement steam l generators.
WCAP-14824, Revision 1, is included as an attachment for reference. WCAP-14824 contains the P/T curves for Byron Unit 1, along with the weld metal data integration for Byron and Braidwood Units 1 and 2 and the Byron /Braidwood fluence methodology justification for ENDF/B-VI cross sections.
2.1 RCS Pressure and Temperature (Pff) Limits (LCO 3.4.9.1/ [ITS-LCO 3.4.3])
1 -
l 2.1.1 The RCS temperature rate-of-change limits defined in Reference 7 are:
I a. A maximum heatup of 100 F in any 1-hour period,
- b. A maximum cooldown of 100*F in any 1-hour period, and c s A maximum temperature change ofless than or equal to 10 F in any 1-hour period j
during inservice hydrostatic and leak testing operations above the heatup and cooldowt
- limit curves.
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-. -.. - _ ~ - - - - - - - - - -
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Operating Limits (Continued)
~ 2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1. The RCS PR limits for cooldown are shown in Figure 2.2 and Table 2.1. These limits are defined in Reference 7. Consistent with the -
methodology described in Reference 1, the RCS PJ limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. In determining compliance with Figures 2.1 and 2.2 and Table 2.1, instrument uncertainties need not be considered since appropriate station operating procedures ensure that the limits contained in the figures and table are not exceeded. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.
The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum permissible temperature in the corresponding P6 curve for heatup and cooldown.
2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.9.3/
[lTS-LCO 3.4.12]).
The power operated relief valves (PORVs) shall each have maximum lift settings ~ in accordance with Figure 2.3 and Table 2.2. These limits are based on References 5 and 13.
The LTOP setpoints are based on the PR limits established by 10 CFR 50, Appendix G without allowance for instrumentation error in accordance with the methodology described -
in Reference 1, the LTOP PORV maximum lift settings shown in Figure 2.3 and Table 2.3 account for appropriate instrument error.
2.3 LTOP Enable Temperature i
The as analyzed LTOP enable temperature is 210*F (Reference 5).
The TS required enable temperature for the PORVs shall be 2 350*F RCS temperature. ;
(Braidwood Unit 2 procedures governing the heatup and cooldown of the RCS require the ;
arming of the LTOP System for RCS temperature of 350*F and below and disarming of LTOP for RCS temperature above 350*F).
Note that the last LTOP PORV segment in Table 2.2 extends to 450 F where the pressure ;
setpoint is 2350 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.
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- BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT *
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Operating Limits (Continued) !
2.4 Reactor Vessel Boltup Temperature (Non-Technical Specification)
The minimum boltup temperature for the Reactor Vessel Flange shall be 2 60 F. Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and j within the RCS vented to atmosphere (Reference 7).
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2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification) l l
1 The minimum temperature at which the Reactor Vessel may be pressurized (i.e., m an unvented condition) shall be 2 60 F, plus an allowance for the uncertainty of the temperature i
instrument, determined using a technique consistent with ISA-S67.04-1994.
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- BRAIDWOOD - UNIT 2 -
PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS i
t LIMITING MATERIAL: WELD METAL LIMITING ART VALUES AT 16 EFPY: 1/4.t. 62.6*F 3/4-t. 55.7'F 4
2500 _ ,,,,,,, ,, , , , , !
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OPERATION I I I L I f 3
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, 1250 / / ACCEPTABLE i 4 / / OPERATION L
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0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.F)
Figure 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100 'F/hr) applicable for the First 16 EFPY* (Without Margins for Instrumentation Errors)
' Applicability date has been reduced to 7.4 EFPY per evaluation in WCAP-14824, App. B, (Reference 2).
4
. l
, BRAIDWOOD - UNIT 2 l
PRESSURE AND TEMPERATURE LIMITS REPORT l MA'ERIAL PROPERTY BASIS LIMITING MA"ERIAL: %TLD METAL 1.tMITING ART VALUES AT 16 EFPY: 1/4-t. 62.6*F 3/4-t. 55.7'F
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2500 ,,,,,,,,,,, , , , , , ,
[ 7898686066868 8 7 i i
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1750 - : UNACCEPTABLE / !
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cJ f 3 1500 /
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/ OPERATION CL. /
y 1000 0 -
a : i;oigo""
a 750 -- rtur.
o ii., ,,
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100 '
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l 250 0
0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.F)
Figure 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) applicable for the First 16 EFPY* (Without Margins for Instrumentation Errors)
\
' Applicability date has been reduced to 7.4 EFPY per evaluation in WCAP-14824, App. B (Reference 2).
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 Braidwood Unit 2 Heatup and Cooldown** Data Points at 16 EFPY*****
(Without Margins for Instrumentation Errors) i 1
Cooldown Curves Heatup Curve
"% Sicad.v State 25 DEG CD 50 DEG CD 100 DEG CD L4nk Test Data 100 DEG Crit. L.imit T P T P T P T P T P T P T P 60 621.00 60 617.85 60 578.29 60 499.51 60 596 98 65 621.00 1% 0.00 174 2000 65 621.00 65 595.85 65 520.34 65 596 98 1% 639.10 1% 2485 70 621.00 70 621.00 70 614.90 70 542.70 75 621.00 70 596.98 196 625.35 75 621.00 75 621.00 75 567.00 75 596.98 1% 613.88 80 621.00 80 621.00 80 621.00 80 593.08 80 $96.98 85 621.00 85 621.00 1% 605.59 85 621.00 85 621.00 85 5%.98 1% 600.02 90 621.00 90 621.00 90 621.00 90 621.00 90 596.98 95 621.00 95 621.00 196 597.31 95 621.00 95 621.00 95 5 %.98 1% 596.98 100 621.00 100 621.r0 100 621.00 100 621.00 105 621.00 100 599.14 1% 599.14 105 621.00 105 621.00 105 621.00 105 603.51 1% 603.51 110 621.00 110 621.00 110 621.00 110 621.00 115 621.00 110 610.11 1% 610.11 115 621.00 115 621.00 115 621.00 115 618.77 196 618.77 120 671.00 a0 621.00 120 621.00 120 621.00 125 641.00 ;?5 621.00 120 623.00 196 629.54 125 621.00 125 621.00 125 621.00 196 642.23 130 621.00 130 621.00 130 621.00 130 621.00 135 621.00 130 621.00 1% 657.17 135 621.00 135 621.00 135 621.00 140 621.00 140 621.00 196 674.19 140 1098.59 140 621.00 196 693.26 140 1096.32 140 693.26 145 ll46.59 196 714.68 150 1198.17 145 714.68 1% 738.24 155 1253.35 150 738.24 196 764.34 160 1312.93 15f 764.34 200 792.79 165 1376.68 160 792.79 205 823.95 170 I444.98 165 823.95 210 857.82 175 1518.49 170 857.82 215 894.51 180 1597.14 175 894.51 220 934.26 185 1681.36 180 934.26 225 977.24 190 1771.61 185 977.24 230 1023.68 195 1868.42 190 1023.68 235 1073.77 200 1971.96 195 1073.77 240 1127.78 205 2082.66 200 1127.78 245 1185.83 210 2201.09 205 1885.83 250 1248.20 215 2327.53 210 1248.20 255 1315.26 220 2462.43 215 1315.26 260 1387.18 220 1387.18 265 1464.42 225 1464.42 270 1547.17 230 1547.17 275 1635.73 235 1635.73 280 1730.69 240 1730.69 285 1832.21 245 1832.21 290 1940 81 250 1940.81 295 2056.75 255 2056.75 300 2180.44 260 2180 44 30$ 2312.46 265 2312.46 310 2453.06 270 2453 06 Applicability date has been reduced to 7.4 EFPY per an evaluation in WCAP-14824, Appendix B (Reference 2).
Heatup and cooldown data include vessel flange requirements of 140*F and 621 psig per 10 CFR 50, Appendix G.
- "For each cooldown rate, the steady-state pressure values shall govern the temperature where no allowable pressure values are provided.
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. . l BRAIDWOOD - UNIT 2 4
PRESSURE ANDTEMPERATURE LIMITS REPORT l
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I 400 0 50 100 150 200 250 300 350 400 AUCTIONEERED LOW RCS TEMPERATURE (DEG. F) i 1
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l Figure 2.3 !
Braidwood Unit 2 Maximum Allowable Nominal PORV Setpoints for the LowTemperature Overpressure Protection (LTOP) System Applicable for the First 16 EFPY*
- Applicability date has been reduced to 7.4 EFPY per evaluation in WCAP-14824, Appendix B (Reference 2).
l 7
. I BRAIDWOOD - UNIT 2 4
PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2 Data Points for Braidwood Unit 2 Maximum Allowable PORV Setpoints for the LTOP System Applicable for the First 16 EFPY*
l PCV-455A PCV-456 '
AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F) (PSIG) RCS TEMP. (DEG. F) (PSIG) 50 497 50 513 70 497 70 513 100 497 100 513 120 452 120 469 170 451 170 468 4
200 614 200 630 250 599 250 '
615 4
300 584 300 600-350 584 350 600 450 2350 450 2359 Note: TO determine maximum allowable lift setpoints for RCS Pressure and RCS Temperatures greater than 350 F, linearly interpolate between the 350'F and 450 F data points shown above.
- Applicability date has been ieduced to 7.4 EFPY per evaluation in WCAP-14824, Appendix B (Reference 2).
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.. . . - . - - . . . . _ . ~ - - -. .. .. .- -
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT '
3.0 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determine changes in material properties. The removal schedule is provided in Table 3.1. The results of these analyses shall be used to update Figures 2.1 and 2.2 and Table 2.1. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely' approaching the withdrawal schedule.
L The pressure vessel material surveillance program (Reference 6) is in compliance with Appendix H to 10 CFR 50," Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTuor, ,
which is determined in accordance with ASTM E208. The empirical relationship between RTNor and the fracture tougimess of the reactor vessel steel is developed in accordance with Appendix G,
" Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM El85-82.
T 9
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' BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Braidwood Unit 2 Capsule Withdrawal Schedule
- Capsule Vessel Location Capsule Lead Removal Time
- Estimated Capsule (Degrees) Factor (EFPY) 2 Fluence (n/cm )
U 58.5* 4.00 1.15 (Removed) 3.933 x 10" X 238.5* 4.02 4.215 (Removed) 1.126 x 10" W 121.5' 4.02 7.97 (EOL Wall) 2.199 x 10"*)
11.95 (1.5 EOL :
Z 301.5* 4.02 Wall (*)) 3.299 x 10" V 61.0* 3.70 Standby --- ;
Y 241.0* 3.70 Standby --- ;
(a) Effective Full Power Years (EFPY) from plant startup.
(b) Maximum end oflicense (32 EFPY) inner vessel wall fluence.
(c) Derived from Table C-4 of WCAP-14824 (Reference 2, which is the Attachment to this report).
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i
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BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.
Table 4.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.
Table 4.2 provides the reactor vessel material properties.
Table 4.3 provides a summary of the Braidwood Unit 2 adjusted reference temperatures (ARTS) at the 1/4T and 3/4T locations for 16 EFPY.
Table 4.4 shows the calculation of ARTS at 16 EFPY for the limiting Braidwood Unit 2 reactor vessel material -Weld Metal WF562 (Based on Surveillance Capsule Data).
. Table 4.5 provides the RTrrs values for Braidwood Unit 2 for 32 EFPY.
Table 4.6 provides the RTyrsvalues for Braidwood Unit 2 for 48 EFPY.
1 I
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BRAIDWOOD - UNIT 2 1
PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 i
Braidwood Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data 4
Fluence
- Material Capsule (n/cm2 , pp(a) Measured FF*ARTwor (FF)2 E>l .0 ARTuor Mev), f l Lower Shell Forging U 3.933 x10'8 0.741 0 0.000 0.550 50D102/50C97
- (Twigential) j X 1.126 x10" 1.033 3 3.099 1.067 i
Lower Shell
- Forging U 3.933 x10'8 0.741 5 3.707 0.550 50D102/50C97 j (Axial) X 1.126x10 l.033 35 36.160 1.067 Sum: 42.996 3.234 Chemistry Factor * = 42.996 + 3.234 = 13.3*F
$ Braidwood 1 i Weld U 3.814x10'8 0.733 10(*) 7.333 0.538 4
. Metal WF562*
! X 1.144x10 l.038 25(*) 25.95 1.077
! Braidwood 2 {
- Weld U 3.933x10'8 0.741 0 0 0.550 Metal WF562(*)
1 X 1.126 x10 l.033 '20(*) 20.66 1.067 l Sum: 53.943 3.232 d
l Chemistry Factor * = 53.943+ 3.232= 16.7"F (a) FF = Fluence Factor = fg2:4:*ios o (b) Braidwood Unit 1 ARTwor values were obtained from WCAP-14243. (Reference 12).
2 (c) Braidwood Unit 2 capsule fluence, FF, and ARTxor values were obtained froni WCAP-14230 1
(Reference 7).
(d) Chemistry Factor = E (FF*ARTwor) / E ((FF) 2),
(e) ARTsor per Ratio Procedure of 10 CFR 50.61 (Reference 10) is not affected, since Ratio = 1.0.
See Table B-4 of WCAP 14824 (Reference 2).
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HRAIDWOOD - UNIT 2 i .
PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2 j
l Braidwood Unit 2 Reactor Vessel Material Properties Chemistry Initial Material Cu (%)(*) Ni (%)(') Factor (') RT sor ( F)*) l Description Closure Head -- -- --
20(*)
20(*)
Lower Shell 0.06 0.77 37.0 -30 Forging Upper Shell 0.03 0.71 20.0 -30 Forging Weld Metal 0.03 0.71 41.0 40 WF562 (a) Chemistry Factors are calculated from Cu and Ni values per Regulatory Guide 1.99, Position 1 (Reference 11).
(b) Initial RTuorvalues are measured, WCAP-14230 (Reference 7).
(c) Closure head and vessel flange Initial RTuorvalues are used for considering flange requirements for the heatup/cooldown curves, WCAP-14230.
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.BRAIDWOOD - UNIT 2 l PRESSURE AND TEMPERATURE LIMITS REPORT l Table 4.3 l
Summary of Braidwood Unit 2 Adjusted Reference Temperatures (ARTS) at the 1/4T and 3/4T Locations for 16 EFPY(*)
16 EFPY(*)
I Material Description 1/4T ART (*F) 3/4T ART (*F) I Lower Shell Forging 35.4 15.3 l Using credible !
surveili m capsule data I (R 2osition 2(*) -6.5*) -13.7*) l l
Upper Shell Forging 5.3 -5.5 (R.G Position 1 )
Weld Metal 112.5 90.2 (RG Position 1(d) 1 Using credible 62.6(b) 55.7*)
surveillance capsule data (RG Position 2(*)
(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Position 1 and Position 2 (Reference 11).
(b) These ART values were used to generate the Braidwood Unit 2 heatup and cooldown curves, WCAP-14230 (Reference 7). l (c) Based on evaluation of WCAP-14230; however applicability date has been reduced to 7.4 EFPY per evaluation in WCAP-14824, Appendix B, Weld Metal Integration for Braidwood Units 1 and 2 (Reference 2).
14
BRAIDWOOD - UNIT 2 1
PRESSURE AND TEMPERATURE LIMITS REPORT
- l i
l' Table 4.4 i
4 Braidwood Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 16 EFPYNat the Limiting Reactor Vessel Material Weld Metal WF562 (Based on Surveillance Capsule ')ata)
Parameter Values
' l Operating Time 16 EFPY" LocationM 1/4T ART 3/4T ART Chemistry Factor, CF (*F) 12.8 118 i
2 Fluence (f), n/cm 6.61 x 10 2.38x10'8 (E>l.0 Mev))*
Fluence Factor, FF 0.884 0.612 l
)
ARTunr= CFxFF(*F) 11.31 7.83 Initial RTuny,,I( F) 40 40 Margin, M(*F) 11.31 7.83 ART = I+(CF*FF)+M, *F 62.6 55.7 per RG 1.99, Revision 2 (a) Fluence, f, is based upon f,,,r(E>1.0 Mev) = 1.100x10 at 16 EFPY, WCAP 14228 (Reference 3).
(b) Applicability date has been reduced to 7.4 EFPY per evaluation in WCAP-14824 Appendix B: Weld Metal Integration for Braidwood Units 1 and 2 (Reference 2).
(c) The Braidwood Unit 2 reactor vessel wall thickness is 8.5 inches at the beltline region.
15
r BRAIDWOOD - UNIT 2 i
PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.5 RTers Values for Braidwood Unit 2 for 32 EFPY
! CF p) pp(b) M RTNor(u) ARTrrs RTpis Material ( F) (*F) ('F) ('F) ('F)
Upper Shell Forging 20.0 2.199 1.214 24.28 -30 24.28 18.6 MK 24-3
! Lower Shell Forging 37.0 2.199 1.214 34.0 -30 44.92 48.9 I Using Surveillance Capsule Data 13.3 2.199 1.214 16.15 -30 16.15 2.3 Circumferential Weld 41.0 2.199 1.214 49.77 40 49.77 139.5 Metal WF562 Using Surveillance 16.7 2.199 1.214 20.27 40 20.27 80.5 Capsule Data")
2 (a) 2.199x10" n/cm (E>1.0 Mev) for 32 EFPY from Braidwood 2 PTS report WCAP-14129 (Reference 8).
i (b) FF (Fluence Factor) = fm2saio nos o (c) Calculated using CF based on surveillance capsule data per RG 1.99, Position 2 (Reference i 1).
l l
16
BRAIDWOOD - UNIT 2
\
PRESSURE AND TEMPERATURE LIMITS REPORT l
Table 4.6 RTvrs Values for Braidwood Unit 2 for 48 EFPY CF go) pp(b) M ARTvrs RTwor(u) RTrrs !
Material ( F) (*F) -(F) ( F) ~(F) 1 Upper Shell Forging 20.0 3.298 1.313 26.26 -30 26.26 22.5 Lower Shell Forging 37.0 3.298 1.313 34.0 -30 48.58 52.6 Using Surveillance 13.3 3.298 1.313 17.0 -30 17.46 4.5 Capsule Data (*)
Circumferential Weld 41.0 3.298 1.313 53.83 40 53.83 147.7 Metal WF562 Using Surveillance 16.7 3.298 1.313 21.93 40 21.93 83.9 j Capsule Data (*)
j 2
(a) 3.298x10" n/cm (E>l.0 Mev) for 48 EFPY from Braidwood 2 PTS report WCAP-14229 (Reference 8).
(b) FF (Fluence Factor) = f<o.2smonog o ;
(c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 11). !
l t
I F
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l 17
BRAIDWOOD - UNIT 2 !
PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References
- 1. Andracheck, J.D., et al,, WCAP-14040-A, " Methodology used to Develop Cold Overpressure / Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 2, January 1996.
- 2. Grendys, P.A., WCAP-14824," Byron Unit 1 Heatup and Cooldown Limit Curves for '
', Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood,"
Revision ~ 1, April 1997.
4
- 3. Peter, P.A., et al., WCAP-14228, " Analysis of Capsule X from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program,"
March 1995.
3
- 4. Terek, E., et al., WCAP-12845, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program," i March 1991.
, 5. Westinghouse Letter to Contmonwealth Edison Company, CCE-96-104, "Braidwood Unit 2 LTOPS Setpoints Based on 16 EFPY P/T Limits," January 24,1996.
- 6. Singer, L.R., WCAP-11188, " Commonwealth Edison Company, Braidwood Station Unit 2 Reactor Vessel Surveillance Program," December 1986.
l
. 7. Peter, P.A., WCAP-14230, "Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," ~ March 1995.
1
- 8. Peter, P.A., WCAP-14229," Evaluation of Pressurized Thermal Shock for Braidwood Unit 2," March 1995.
i
- 9. Lippencott, E.P. WCAP-14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation," April 1994.
l
- 10. 10 CFR 50.61," Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,"(PTS Rule) January 18,1996.
I1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, " Radiation ,
Embrittlement of Reactor Vessel Materials," Revision 2, May 1988. )
- 12. Peter, P.A., WCAP-14243,"Braidwood Unit i Heatup and Cooldown Limit Curves for Normal Operation," March 1995.
18
BRAIDWOOD - UNIT 2
- 1 t
PRESSURE AND TEMPERATURE LIMITS REPORT l References (Continued) 13.
Comed Calculation BRW-96-9061/BYR 96-293," Channel Accuracy for Power j-Operated Relief Valve (PORV) Setpoints and Wide Range RCS Temperature !
] Indication (Unit 2 Original Steam Generators)," Revision 0.
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I 19 '
T- .
s WCAP-14824, REVISION 1 J
BYRON UNIT 1 HEATUP AND COOLDOWN LIMIT CURVES FOR I NORMAL OPERATION AND SURVELLANCE WELD METAL INTEGRATION FOR BYRON AND BRAIDWOOD 4
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N l WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14824, Revision 1 i Byron Unit 1 Heatup and Cooldown Limit Curves l For Normal Operation
- and Surveillance Weld Metal integration i for Byron and Braidwood A
I l P. A. Grendys April 1997 1 Work Performed Under Shop Order CPEP-139 Prepared by the Westinghouse Electric Corporation for the Commonwealth Edison Company Approved: d4k W l C. H. Boyd, Manager 0~ N ; Engineering & Materials Technology WESTINGHOUSE ELECTRIC CORPORATION Nuclear Services Division P.O. Box 355 Pittsbugh, Pennsylvania 15230-0355 01997 Westinghouse Electne Crogetion All Rights Reserved
i PREFACE This report has been technically reviewed and verified by:
/
2" _c ' 8 T. J. Laubham
^
" l
/ L/
i i
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l l 1 .f l i : b l I
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l l 1 l l Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 1
ii 1 - b TABLE OF CONTENTS 3 LI ST OF FI G U R ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . li 4 LI ST OF TABLE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv l 3 e !, INTRODU CTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 j 1 1 2 FRACTURE TOUGHNESS PROPERTIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 l {
- 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS . . 3 i
4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE . . . . . . . . . . . . . . 6 ). 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES . . . . 13 i 6 R EFER ENC ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 i ! i APPENDIX A - WELD METAL INTEGRATION FOR BYRON UNITS 1 AND 2 . . . . . . . . . A-0 i i i APPENDIX B - WELD METAL INTEGRATION FOR BRAIDWOOD UNITS 1 AND 2 . . . . B-0 i i APPENDIX C - BYRON /BRAIDWOOD FLUENCE METHODOLOGY JUSTIFICATION I AND TIME-DEPENDENT CAPSULE FLUENCE VALUES . . . . . . . . . . . . C-0 i i i i
'I 4
1 ? 4 i i April 1997
, e Byron Unit 1 Heatup and Cooldown Umit Curves
iii LIST OF FIGURES 1- Byron Unit 1 Reactor Coolant System Heatup Umitations (Heatup Rates up ; to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for instrumentation Errors) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2 Byron Unit 1 Reactor Coolant System Cooldown Umitations (Cooldown Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3' Byron Unit 1 Reactor Coolant System Heatup Umitations (Heatup Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for J Instrumentation Errors; Margin of 74 psig for Pressure Difference Between ! Pressure instrumentation and the ' Reactor Vessel Beltline Region) . . . . . . . . . . . 17 l 1 4 Byron Unit 1 Reactor Coolant System Cooldown Umitations (Cooldown Rates , up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for l Instrumentation Errors; Margin of 74 psig for Pressure Difference Between i Pressure Instrumentation and the Reactor Vessel Beltline Region) . . . . . . . . . . . 18 i l Byron Unit 1 Heatup and Cooldown Umit Curves April 1997
iv i. - LIST OF TABLES 1 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1 Ba se Mate rials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 . 2 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1 - Weld Material (Using Byron 1 & 2 Chemistry Test Results) . .. . . . . . . . . . . . . . . . . 8 3 Byron Unit 1 Reactor Vessel Material Properties . . . . . . . . . . . . . . . . . . . . . . . . . 9 i . 4 Calculation of Chemistry Factors Using Credible Byron Units 1 and 2 Surveillance Capsule Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
- \
- 5 Calculation of Adjusted Reference Temperatures (ART) at 12 EFPY for the ;
Limiting Byron Unit 1 Reactor Vessel Material - Intermediate Shell Forging l SP-5933 (based on credits sunmillance capsule data) . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 1 1 6 Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T f
- Locations for 12 EFPY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 l 7 Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins for Instrumentation Errors 1
includes 1) Vessel flange toqurements of 180*F and 621 poig per 10CFR50. ............. 19 8 Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins for Instrumentation Errors includes 1) Vessel flange requirements of 180*F and 621 peig per 10CFR50, and 2) Pressure i adjustment of 74 poig to account for possure difference between the wide-range pressure transmitter and the irrutng belthne region of the reactor vessel. . . . . . . . . . . . . . . . . . . . . . . . 20 4 v l I 4 l April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves ; I
1 4 !~ ? j 1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTuor (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTuor of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTuo7, and adding a margin. The unirradiated RTuor is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F. RTuo7i ncreases as the materialis exposed to fast-neutron radiation. Therefore, to find the most limiting RTuoy at any time period in the reactor's life, ARTuo, due to the radiation exposure associated with that time period must be added to the unirradiated RT u or(muoy). The extent of the shift in RTuor is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, " Radiation Embnttlement of Reactor Vessel Materials"N, Hegulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTuoy + ARTuor + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad / base metal interface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves. l d April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves
, 2 i l i t 2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure f boundary are determined in accordance with the NRC Regulatory Standard Review Plan #1 The pre-irradiation fracture-toughness properties of the Byron Unit 1 reactor vessel are presented in Table 3. The post-irradiation fracture toughness properties of the reactor vessel beltline material were obtained directly from the Byron Unit 1 Reactor Vessel Radiation , Surveillance ProgramDl. Credible surveillance data is available for two capsules (Capsules U and X) for Byron Unit 1. .This capsule data is used to calculate chemistry factors (See Table I
- 4) in addition to those calculated per Regulatory Guide 1.99, Revision 2.
Additionally, per the request of the Commonwealth Edison Company, the survei! lance weld data from the Byron Unit 1 and Byron Unit 2 surveillance programsHI has been integrated pursuant to 10 CFR 50.61 in accordance with Regulatory Guide 1.99, Revision 2, Position 2. In addition to the credible surveillance weld data from Byron Unit 1, credible surveillance weld data is available for two capsules (Capsules U and W) for Byron Unit 2. The chemistry factor values resulting from the weld metal integration for Byron Units 1 and 2 are presented in Section 4 of this report. See Tables 1 through 4. l l A complete technical justification for the Byron Units 1 and 2 weld metal integration is presented in Appendix A of this report. I Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
3
~
4 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE l RELATIONSHIPS , Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements **l specifies fracture toughness requirements for ferritic materials of pressure retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins
- of safety during any condition of normal operation, including anticipated operational 1 occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements. Section XI, Division 1, " Rules for Inservice inspection of Nuclear Power Piant Components **l, Vessels, contain ihe conservative methods of analysis.
The ASME approach for calculating the allowable limit curves for various hcatup and cooldown rates specifies that the total stress intensity factor, K,, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the referenco stress intensity factor, K , for the metal temperature at that time. K, is obtained from the reference fracture toughness curve, defined in Appendix G of the ASME Code, Section Xi'). t The K, curve is given by the following equation: K,=26.78+1.233 e where, K, = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT, i Therefore, the goveming equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: C.KgK,<K, (2) where, K. = stress intensity factor caused by membrane (pressure) stress K, = stress intensity factor caused by the thermal gradients K, = function of temperature relative to the RT, of the material C= 2.0 for Level A and Level B service limits C= 1.5 for hydrostatic arid leak test conditions during which the reactor core is not critical Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
~
4 l i i. + j At any time during the heatup or cooldown transient, K, is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value , for RT., and the reference fracture toughness curve. The thermal stresses resulting from - the temperature gradients through the vessel wall are ca'culated and then the corresponding (thermal) stress intensity factors, K , for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are l calculated. 1 For the calculation of the allowable pressure versus coolant temperature during cooldown, the 8 reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the i l vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the l wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves , are constructed for each cooldown rate of interest. I j The use of the composite curve in the cooldown analysis is necessary because control of the j l cooldown procedure is based on the measurement of reactor coolant temperature, whereas i 3 the limiting pressure is actually dependent on the material temperature at the tip of the I assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of K, at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K, exceeds K., the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure temperature relationships are inW for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by intamal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K, for the 1/4T crack during heatup is lower than the K, for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower Aprl! 1997 Byron Unit 1 Heatup and Cooldown IJmit Curves
5 l K, values do not offset each other, and the pressure temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when l the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for l steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concems the calculation of the l pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location I from tha outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal l ! stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point ay-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken l from the curves under consideration. The use of the composite curve is necessary to set l conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling cor,Ger, switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion. 10 CFR Part 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange regens. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RT by at least 120*F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is 621 psig for Byron Unit 1. The limiting unirradiated RT,,,, of 60*F occurs in the closure head flange of the Byron Unit 1 reactor vessel, so the minimum allowable temperature of this region is 180*F at pressures greater than 621 psig. This limit is shown in Figures 1 through 4 wherever applicable. l I s April 1997 Byron Unit 1 Hestup and Cooldown Limit Curves
i .
- t. 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE 4
1 From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each f material in the beltline region is given by the following expression: i 1 i ART =ItnnIRTgARTgMarpin (3) ! Initial RT, is the reference temperature for the unirradiated material as defined in paragraph W l NB-2331 of Section til of the ASME Boiler and Pressure Vessel Cc:le . If measured values j of initial RT, for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and l standard deviation for the class. 3 i ART is the mean value of the adjustment in reference temperature caused by irradiation j
- - and should be calculated as follows
s A RT,= CF. f** (4) To calculate ART, at any dspth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth. f =f .e N (5) where x inches (vessel beltline thickness is 8.5 inchest ") is the depth into the vessel wall measured from the vessel clad / base metal interface. The resultant fluence is then placed in Equation 4 to calculate the ART, at the specific depth. The calculated surface fluence for Byron Unit 1 upper and lower shel! forgings and circumferential weld at 12 EFPY is 8.10 x 10" n/cm'. This fluence value was calculated
- rom the surveillance Capsule X analysis presented in WCAP-13880R.
CF (*F) is the chemistry factor, obtained from the tables in Reference 1, using the average values of copper and nickel content as calculated in Tables 1 and 2 and reported in Table 3. The chemistry factors were also calculated using the surveillance capsule data in Table 4. The Ratio Procedure, as documented in paragraph (c)(2)(ii)(B) of 10 CFR Part 50.61, was used to adjust the measured values of ART, for differences in copper and nickel content by multiplying them by the ratio of the chemistry factor for the vessel material (best-estimate chemistry) to that for the surveillance weld. April 1997 Byron Unit 1 Heatup and Cooldown Umit Curves
I 7 J All materials in the beltline region of Byron Unit 1 reactor vessel were considered in l determining the limiting material. Sample calculations to determine the ART values for the Weld Metal at 12 EFPY are shown in Table 5. The resulting ART values for all beltline region 4 materials at the 1/4T and 3/4T locations are summarized in Table 6, where it can be seen that d the limiting material is the intermedia'.a Shell Forging SP 5933 (based on credible surveillance capsule data). The 1/4T and 3/4T ART values for intermediate Shell Forging SP 5933 (based on credible surveillance capsule data) will be used in the generation of heatup and cooldown j curves applicable to 12 EFPY. TABLE 1 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1 Base Materials i Reference intermediate Shell Forging Lower Shell Forging SP 5933 SP 5951 Byron Unit 1 0.034 0.73 0.04 0.64 l HU/CD Limit ON 0.032 0.791 0.03 0.75 Mer RW FDRT/ SRPLO 009(94) 0.05 0.73 0.036 0.735 January 1994 Average 0.0364 0.747 0.04 0.64 Standard Deviatkm 0.007 0.023 0 0 Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
f
- 8 j TABLE 2 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1
- Weld Material (Using Byron 1 & 2 Chemistry Test Results) i i
Best Estimate Reference Qg Hi l BAW-2261 0.024 0.7 d B&W Weld Qualification ! 0.031 0.46 B&W Wold Qualification 0.03 0.72 B&W Wald Qualification . ' 0.068 0.48 B&W Weld Qualification 0.114 0.54 i B&W Weld Qualification 0.148 0.6 l B&W Wold Qualification 0.053 0.62 BaW Wold Qualification 0.059 0.62 B&W Weld Qualification Surv. CF = 27 Byron 1 Surveillance Data See Below 0.022 0.690 -> 0.02 0.69 0.023 0.712 -> 0.02 0.71 Surv. CF = 27 j Byron 2 Surveillance Data See Below 0.057 0.614 -> 0.06 0.61 Best Est. CF = B2 Best Estirnate Chemistry: 0.043 0.095 Byron 1 & 2 Ratio = 3.0 i Standard Deviation: i i SurveiNance Data Cherrustry Results: Byron Unit 2 i Reference Q g l Syron (Araft 7 0.65 g g WCAP 10398W 0.03 ! Reference 0.026 0.71 WCAP 1243184 0.024 0.740 WCAP-951783 0.67 0.024 0.786 WCAP-1165153 0.023 l 0.022 0.665 0.022 0.704 0.021 0.714 0.020 0.631 0.021 0.706 8 0.021 0.741 ' O.022 0.713 0.020 0.697 0.021 0.714 0.019 0.668 0.020 0.704 0.022 0.759 0.021 0.714 0.020 0.694 0.020 0.706 0.020 0.678 0.021 0.677 0.020 0.695 0.023 0.677 0.019 0.689 0.021 0.744 0.021 0.680 0.021 0.680 0.022 0.738 0.021 0.667 0.022 0.771 0.024 0.677 WCAP-14064'"3 0.024 0.705 0.022 0.807 0.023 0.706 0.021 0.834 0.023 0.898 0.882 0.024 0.996 WCAP-13880M 0.024 0.023 0.711 0.022 0.678 0.024 0.708 93 93 0.024 0.716 Average 0.022 0.000 0.024 0.715 0.024 0.707 0.024 0.720 0.024 0.717 0.024 0.711 0.024 0.706 0.024 0.707 9M 9112 Average 0.033 0.712 April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves J
9 TABLE 2 NOTES: (a) The weld material in the Byron Unit 1 surveillance program was made of the same wire and flux as the reactor vessel intermediate to lower shell girth seam weld. (Weld seam WF-336 Wire Heat No. 442002, Flux Type Linde 80, Flux Lot No. 8873) (b) The E, a Unit 2 surveillance weld is identical to that used in the reactor vessel core region girth seam (WF-447). The weld wire is type Linde MnMoNi (Low Cu-P), heat number 442002, with a Linde 80 type flux, tot number 8064. TABLE 3 Byron Unit 1 Reactor Vessel Material Properties Material Description Cu (%) Ni(%) Chemistry initial Factor) RTuor("F)*) Closure Head Flange - 0.74 - 60(*) Vessel Flange - 0.73 - 10) Intermediate Shell Forging 0.0364 0.747 23.8 40 SP-5933 Lower Shell Forging SP-5951 0.04 0.64 26.0 10 Circumferential Weld WF-336 0.06 0.61 82.0 -30 NOTES: (a) Chemistry Factors are calculated from Cu and Ni values per Regulatory Guide 1.99, Revision 2. (b) Initial RT,e, values are measured values. (c) Closure head and vessel flange initial RTeer values are used for considering flange requirementd51 for the heatup/cooldown curves. Byron Unit 1 Heatup and Cooldown Lime Curves April 1997
10 TABLE 4 Calculation of Chernistry Factors Using Credible Byron Units 1 and 2 Surveillance Capsule Data 1 Material Capsule Capsule FP*) Meas. FP FF' Fluence f A RT, ART ,
. inter. Shell U 3.72x10 O.727 0 0 0.529 Forging SP-5933 (Tangential) 'X 1.39x10 1.091 30 32.73 1.19 ,
- l Inter. Shell U 3.72x10'8 0.727 0 0 0.529 Forging SP-5933 X 1.39x10 1.091 30 32.73 1.19
;,,)
Sum: 65.46 3.44
. Chemistry Factor'* = 65.46 + 3.44 = 19.0*F l Byron 1 Weld U 3.72x10 O.727 0 0 0.00 0.529 1 Metal WF 336*)
X 1.39x10 1.091 35 10540 114.56 1.19 1 l Byron 2 Weld U 3.996x10 O.746 0 0 0.00 0.557 Metal WF-447(* W 1.211x10 1.053 30 90'" 94.77 1.110 Sum: 209.33 3.386 Chemistry Factod' = 209.33 + 3.386 = 61.8'F ~ NOTES. (a) FF = Fluence Factor - fu e.e men (b) Byron Unit 1 ART, values were obtained from the surve!"ance Capsule X analysis (WCAP-13880). The Byron Unit 1 capsule fluence values were recalculated using the ENDF/B-V scattering . cross sections in 1994 and are documented in WCAP 14044"". (c) Byron Unit 2 capsule fluence, FF, and ART, values were obtained from the surveillance Capsule W analysis (WCAP-14064"") using the ENDF/B-V scattering cross sections. (d) Chemistry Factor = MFPART ) + hFF')
- (e) Adjusted ART, per Ratio Procedure of 10 CFB 2.61. Ratio = 3.0. See Table 2.
i l Byron Unit 1 Heatup and Cooldown Limit Curtes April 1997
I 11
) ;l Margin is calculated as, M = 2 ho,8 + o,8 The standard deviation for the initial uo RT y margin .
term, o,, is O'F when the initial RTuor is a measured value, and 17'F when a generic value is I available. The standard deviation for the ARTuo, margin term, o,, is 17'F for the plate, and 8.5*F for the plate (half the value) when surveillance data is used. For welds, o, is equal te ! 28'F when survei' lance capsule is not used, and equal to 14*F when credible surveillance capsule data is used. o, need not exceed 0.5 times the mean value of ART uor. TABLE 5 Calculation of Adjusted Reference Temperatures (ART) at 12 EFPY for the Limiting Byron Unit 1 Reactor Vessel Material intermediate Shell Forging 5P 5933 (based on credible surveillance capsule data) I Parameter Values . Operating Time 12 EFPY Material Intermediate Shell Forging SP-5933 Location 1/4T 3/4T i l Chemistry Factor, CF (*F) 19.0 19.0 Fluence, f (10" n/cm') O.486 0.175 Fluence Factor, FF 0.799 0.538
- l ARTuor = CF x ff (*F) 15.2 102 initial RTuct, I (*F) 40 40
- Margin, M (*F) 15.2 10.2 Adjusted Referenes Temperature (ART), (*F) 70 60 per Regulatory Guide 1.99, Revision 2 i
NOTES: (a) The Byron Unit 1 reactor vessel wall thickness is 8.5 ircies at the beltline region. (b) Fluence, f, is based upon fu (10" n/cm', E>1.0 MeV) = 0.810 at 12 EFPY. 1 i Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
1 12 l TABLE 6 Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T Locations for 12 EFPY Material 12 EFPY 1/4T ART 3/4T ART Intermediate Shell Forging 78 66 SP-5933 (RG Posthon 1*') s using credible surveillance 70*' 60*' capsule data (RG Posibon 2") 1 Lower Shell Forging SP-5951 52 38 (RG Position 1")) Circumferential Weld WF-336 92 58 (RG Position 1*') i using credible surveillance 47 31
- capsule data (RG Position 2"')
4 l NOTES: (a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Revision 2, Positions 1 and 2. (b) These ART values were used to generate the Byron Unit 1 heatup and cooldown curves. l Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
i : ' 13 j 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT ! CURVES 4 Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor l 4 coolant system have been calculated for the pressure and temperature in the reactor vessel ' beltline region using the methods"'l discussed in Section 3 and 4 of this report. This approved methodology.is also presented in WCAP-14040-NP-A"81, dated January 1996. l Since indication of reactor vessel beltline pressure is not available on the plant, the pressure l difference between the vnde-range pressure transmitter and the limiting beltline region must j I be accounted for when using pressure-temperature limit curves presented in Figures 1 and 2. Generic calculations (based upon four active loops and one operating RHR pump) have l determined that the pressure indicated by the reactor coolant system wide-range instrumentation should be assumed to be 74 psig less than that at the reactor vessel beltline l for Byron Unit 1"51. Figures 3 and 4 do include this pressure difference of 74 psig. Figures 1 and 3 present the heatup curves without margins for instrumentation errors using a heatup rate of 100*F/h'r applicable for the first 12 EFPY. Figure 3 2 and 4 present the cooldown curves without margins for instrumentation errors using cooldown rates up to 100*F/hr applicable for the first 12 EFPY. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 through 4. This is in addition to other criteria which must be met before the reactor is made critical. The reactor must not be made critical until pressure-temperature combinations are to the right of the enticality limit line shown in Figures 1 through 4. The straight-line portion of the cnbcality limit is at the minimum permissible temperature for the 2485 psig inservice
)
hydrostatic test as required by Appendix G to 10 CFR Part 50. The goveming equation for the hydrostatic test is defined in Appendix G to Sechon XI of the ASME Code as follows: (6) i 1.5K,<K, where, K. is the stress intensity factor covered by membrane (pressure) stress, K.= 26.78 + 1.233 e #N'" * *1, T is the minimum permissible metal temperature, and RT, is the metal reference nil-ductility temperature The criticality limit curve specifies pressure-temperature limits for core operation to provide addibonal margin during actual power production as specified in Reference The '5. April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves
' 14 i pressure-temperature limits or core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and - cooldown calculated as described in Section 3 of this report. The minimum temperature for the inservice hydrostatic leak tests for the Byron Unit 1 reactor vessel at 12 EFPY is 203*F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel. 9
- Figures 1 through 4 define all of the above limits for ensuring prevention of nonductile failure for the Byron Unit 1 reactor vessel. The data points used for the heatup and cooldown l
- pressure-temperature limit curves shown in Figures 1 through 4 are presented in Tables 6 and
- 7.
Additionally, Westinghouse Engineering has reviewed the minimum boltup temperature j requirements for the Byron Unit 1 reactor pressure vessel. According to Paragraph G-2222 of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Appendix G, the reactor vessel may be bolted up and pressurized to 20 percent of the initial hydrostatic test pressure at the initial RTug of the material stressed by the boltup. Therefore, since the most limiting initial RTum value is 60*F (closure head flange), the reactor vessel can be bolted up at this temperature. I April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves
l
'" 15 1
i MATERIAL PROPERTY BASIS
\
LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP-5933 (using surv. capsule data) l LIMITING ART VALUES AT 12 EFPY: 1/4T, 70*F ! 3/4T, 60*F ! 1 2500 i, , , , , , ,, Tssssseosass i' ' t i il ' ;l ' l i .J i 'l - il
~7 e i ,i. .
i !
./ i f i f i i e ii' i e '
2250 i l l
,f . f, , ,_
m T LEAI' TEST LIMIT Ci i i IIl ' l' > T ! i/ .'I' >
/ii i ,
/.
tao i , , , i ill . ; r . t ,,, , , i 2000 ' ' '
'l
/,
W , . .
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j g . * '
* .! l '
1500 s i, , .' . 1. iy g i i! ! >i !1 . i ! , .Mi /'! i - - i : j
!i i i i! fi/ l i /i i i i . . . ,
i
,'l !EEATUr EATE / jf ' [!;, ! '
ca 1250 , , , or To too r/ar. g,,, , , , , f ' ,' cc' 6 !t i .
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i OPERATION i.i ! g y.!! ! I I! e iii ,/ i ! i i/ t
' ii i ! ! !!i; is !/ ' 1 Iif ' l ii_i . I 1000 .' ..,
y i > f i.fi ,, i ! , , ,, , ,, , , 4 6 ti ! I i i /l iV I i! i i ! ! '
! ! ' i ! i i i C i i i .._ ! i t ! IA ii i i i ! ! ! Ii i i 1' a i i
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t t .* 750 ' '
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ee ,ii ie ii i ! i i i i i i , . . , U iii iti ii i >
! . i i i i , nii
.-=
! I I I i iiI,
! I i ii !It ! ii t i i t i i i i i i 1 i i i i i i e 500 ' ' '
i >> i! t i i . : ! , ! i ! ! ! i . i g:: i i6 !,i k' t l I l i!*I i ; !l !. I ! 1l1 .
~
I N m i i ; ,- i i i i i !i h i!'!! Y' MINIMUM BOLTUF I. II ' !I' ' 250 ' l,l TEMP. AT 60*F cantica'Livy Lawit sassa on , i ! ii i Isssatics urssosta1'ac tsst ! -
,j;j 1 i ! Tsurasatuas (ses r . ros ts: I! i
, ! i ; i sant s cs pas s es or 1'o as.e arrY , ,j 1 i.. i i ie . ,.
0 , i i i # l 0 50 100 150 200 250 300 3$0 4d0 4$0 500 Indicated Temperature (Deg.F) FIGURE 1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors) Byron Unit 1 Heatup and Cooldown Limit Curves - April 1997
16 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP 5933 (using surv. capsule data) LIMITING ART VALUES AT 12 EFPY: 1/4T, 70*F 3/4T, 60*F 2500 i , , _ sassasseessee ~~
.f i ~
+ i i i ! . i !/ i i i 2250 '; ' '
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! i i i e i e i, t 1 i iN MINIMUM BOLTUP 8 I i i !I i ! !I i i i i i i i !' TEMP. AT 60*F tii ii!l -
I . Ii e i i! i l I ii 1 i i i i i !! l ! l! t 0 , , iii l , ! I i i i i 0 50 100 150 2b0 250 300 350 400 450 500 Indicated Temperature (Deg.F) FIGURE 2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for instrumentation Errors) Byron Unit 1 Heatup and Cooldown Urnit Curves April 1997
17 4 i MATERIAL PROPEFTY BASIS e LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP-5933 (ueing surv. capsule data) ! . L!MITING ART VALUES AT 12 EFPY: 1/4T, 70*F
- 3/4T, 60*F i 2500 , , . ,
1
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__. usesstses svenostaPac test l, i ; TEMP. AT GO'F jjU!! 'ja!:N'Or#bo'!!.s'!rrY j
, , , , i 0 is i , i 0 50 160 1$0 200 2$0 3d0 3$0 400 4$0 500 Indicated Temperature (Deg.F)
FIGURE 3 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for instrumentation Enors; Margin of 74 peig for Pressure Difference Between Pressure instrumentation and the Reactor Vessel Beltline Region) Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 -
18 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 5P-5933 (using surv. capsule data) LIMITING ART VALUES AT 12 EFPY: 1/4T, 70*F 3/4T, 60*F 2500 , i _ .............l. -- ~ a, - -- i , . , i i i,: . 2250 , .
/' ,
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- 7' ' ' '
I l g 36 s/ I e ! l > l i- i a g l 60 fi i li,1 i l
! I I- ! !
i I i i 7 100 i4 j; } i i t t ! i t j i i
~ i i ; } i j T l i ii I ie i i i i i i -
i i iii ! 1 ii ii 250 4 i ii m N'{ i i i i , , iii iii i i 1 , i e i l ! i 'I ! i i! !! ( i i i ii N MINIMUM BOLTUP ! I i i ' i I ' ' t I i i <l ll TEMP. AT 60*F l ', ! l j l ! lll i , i i iii, i 0
, , i i . iii i ! ii 0 5'O 100 150 2b0 250 300 350 400 450 500 Indicated Temperature (Deg.F)
FIGURE 4 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors; Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and the Reactor Vessel Beltline Region) Byron Unit 1 Hertup and Cooldown Lirrdt Curves April 1997
i. 19 h TABLE 7
; Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins
, for Instrumentation Errors t ! includes 1) Vessel flange requirements of 180*F and 621 psig per 10CFR50. ; Cooldown Curves Heatup Curve i Steady State 25F 50F 100F 100F Criticality. Umit Leak Test Umit - T P T P T P T P T P T P T P 3 60 621 60 595 60 554 60 470 60 821 203 0 182 2000 ! 65 621 65 610 65 570 65 489 65 621 203 0 203 2485 70 621- 70 621 70 587 70 509 70 621 203 0 75 621 75 621 75 605 75 531 75 621 203 0 80 621 80 621 80 621 80 554 80 621 203 671 85 621 85 621 85 621 85 579 85 621 203 657 90 621 90 621 30 621 90 607 90 621 203 646 . l
! 95 621 95 621 95 621 95 621 95 621 203 639 i
! 100 621 100 621 100 621 100 621 100 621 203 634 l j 105 621 105 621 105 621 105 621 105 621 203 632 l j 110 621 110 621 110 621 110 621 110 621 203 633
- 115 621 115 621 115 621 115 621 115 621 203 637 j 120 621 120 62# 120 621 120 621 120 621 203 642 125 621 125 6T.1 125 621 125 621 125 621 203 651 j 130 621 130 (21 130 621 130 621 130 621 203 661 135 621 135 621 135 621 135 621 135 621 203 674 l 140 621 140 621 140 621 140 621 140 621 203 689 l l 145 621 145 621 145 621 145 621 203 707 I
- 150 621 150 -621 150 621 203 727 l 155 621 155 621 203 749
- 160 621 160 621 203 774 165 621 165 621 205 801 170 621 170 621 210 831 l 175 621 175 621 215 864 l 100 621 180 621 220 900 180 1483 180 900 225 938 185 1559 185 936 230 980 i 190 1640 190 900 235 1026 8
195 1728 195 1026 240 1075
- 200 1821 203 1075 245 1128 205 1921 205 1128 250 1186 l 210 2029 210 1186 255 1247
. 215 2143 215 1247 260 1313 220 2266 220 1313 265 1385 4 225 2397 225 1385 270 1461 230 1461 275 1543 i 235 1543 280 1630
- 240 1630 285 1724 i 245 1724 290 1825 250 1825 295 1933
, 255 1933 300 2048 200 2048 305 2171 j 265 2171 310 2302
- 270 2302 315 2441
! 275 2441 4
' (Con 6gurabon #930331ssuuutan for Cooldown, #2756858809292 for Heatup) j Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
20 e [ TABLE 3
- Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins i for instrumentation Errors j i i includes 1) Vessel flange requirements of 180*F and 621 psig per 10CFR50, and 2) Pressure adjustment of I l
3 74 psig to account te pressure difference between the wide range pressure transmitter and the limiting beltline region of the reactor vessel. l l e-Cooldown Curves , Heatup Curve } 100F 100F Criticality. Umit Leak Test Umit Steady State -25F 50F T T T i T- P T P T P T P P P P 1 60 .547 60 521 60 480 60 396 60 547 203 0 182 2000 j 65 547 65 536 65 496 65 415 65 547 203 0 203 2485 } 70 547 70 547 70 513 70 435 70 547 203 0 l 75 ' 547 75 547 75 531 75 457 75 547 203 0 80 547 80 547 80 547 80 480 80 547 203 597 l 45 547 85 547 85 547 85 505 85 547 203 583
- 90 547 90 547 90 547 90 533 90 547 203 572 f 95 547 95 547 95 447 95 547. 95 547 203 565 4 100 547 100 547 100 547 100 547 100 547 203 560 l 105 547 105 547 105 547 105 547 105 547 203 558 4 110 547 110 547 110 547 110 547 110 547 203 559 l 115 547 115 547 115 547 115 547 115 547 203 563 i 120 547 120 547 120 547 120 547 120 547 203 568 s 125 547 125 547 125 547 125 547 125 547 203 577 1
130 547 130 547 130 547 130 547 130 547 203 587 133 547 135 547 135 547 135 547 135 547 203 600 140 547 140 547 140 547 140 547 140 547 203 615 145 547 145 547 145 547 145 547 203 633 150 547 150 547 150 547 203 653 155 547 155 547 203 675 160 547 160 547 203 700 165 547 165 547 205 727 170 547 170 547 210 757 o 175 547 175 547 .215 790 180 547 180 547 220 826 ; 180 1400 180 826 225 864 185 1485 185 864 230 906 l 190 1566 190 906 235 952 1 195 1654 195 952 240 1001 200 1747 200 1001 245 1054 205- 1647 205 1054 250 1112 210 1955 210 1M2 255 1173 215 2000 215 1173 260 1239 220 2192 220 1239 265 1311 j 225 2323 225 1311 270 1387 1 230 1387 275 1469 235 1469 200 1556 240 1556 285 1850 245 1950 290 1751 250 1751 295 1859 255 1859 300 1974 200 1974 305 2007 265 2007 310 2228 270 2228 315 2367 (Configurabon #93048Wtatan003 for Coot %wn, #9291115665880 for Heatup) Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
- 21 i !, 6 REFERENCES , t 1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel
- Materials', U.S. Nuclear Regulatory Commission, May,1988.
2 Fracture Toughness Requirements", Branch Technical Position MTER 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981. i i 3 WCAP 9517, " Commonwealth Edison Co. Byron Station Unit 1 Reactor Vessel l Radiation Surveillance Program', J. A. Davidson, July 1979. 4 WCAP-10398, " Commonwealth Edison Co. Byron Station Unit 2 Reactor Vessel i Radiation Surveillance Program', L R. Singer, December 1983. I 5 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", Federal Register, 1 ! Volume 60, No. 243, dated December 19,1995. ; t ! 6 1992 Section X!, Division 1, of the ASME Boiler and Pressure Vessel Code, Appendix ! j G, " Vessels'. ! 7 1989 ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Appendix G, i " Fracture Toughness Criteria for Protection Against Failure". i 8 1989 Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph l i NB-2331, " Material for Vessels". 9 WCAP-13880, " Analysis of Capsule X from the Commonwealth Edison Company Byron 4 Unit 1 Reactor Vessel Radiation Surveillance Program', P. A. Peter, et al., January ! 1994. ; ! 10 WCAP-14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation", E. ! i P. Lippincott, April 1994. i 3 11 WCAP-14064, " Analysis of Captule W from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program", P. A. Peter, et al., November 4 1994. I 12 WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves", W. S. Hazelton, et al., April 1975. Byron Unit 1 Heatup and Cooldown Uml: Curves April 1997
22 i 13 WCAP-14040-NP-A, Revision 2, " Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D. Andrachek, et al., January 1996. i 14 Babcock & Wilcox drawing numbers 184557E, Rev. 2; 185266E, Rev. 2; 185297E. Rev. 2; 1853C. Rev. 2; " Reactor Vessel Longitudinal Section'. 15 Nuclear Safety Advisory Letter, NSAL-93-005A, " Cold Overpressure Mitigation System-(COMS) Nonconservatism", L R. Hardwick and H. A. Sepp,3/10/93. 16 WCAP-14063, " Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation", P. A. Peter, November 1994. 17 WCAP-13881, " Evaluation of Pressurized Thermal Shock for Byron Unit 1", P. A. Peter, January 1994. 18 WCAP-14054, " Evaluation of Pressurized Thermal Shock for Byron Unit 2', P. A. Peter, August 1995. 19 WCAP-14242, " Evaluation of Pressurized Thermal Shock for Braidwood Unit 1*, P. A. Peter, March 1995. 20 WCAP-14229, " Evaluation of Pressurized Thermal Shock for Braidwood Unit 2", P. A. Peter, March 1995. 21 WCAP-11651, " Analysis of Capsule U From The Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko, et al., November 1987. 22 WCAP-12431, " Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Survel!iance Program", E. Terek, et al., October 1989. l 1 J Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
d A-0 I 1 s J I 4 APPENDIX A
- WELD METAL INTEGRATION FOR BYRON UNITS 1 AND 2 a
i k 4 i 4 1 l l .i 1 4 i 4 Byron Unit 1 Heatup and Cooldown Limit Cunes April 1997
A-1 l
- INTRODUCTION
4 Westinghouse performed an evaluation to determine if the weld wire data of the Byron Units 1 l and 2 surveillance programs can be integrated. The evaluation was based on the following l I criteria:
- 1. What weld wire heat number, flux, and flux lot were used to fabricate the surveillance l
program weld metal of each unit,
- 2. What vendor fabricated the welds and in what time frame.
4
- 3. What heat treatment did each surveillanm program weld receive,
! 4. Is the initial RT, of the welds the same or relatively close,
- 5. Is the initial upper shelf energy of the welds the same or relatively close,
- 6. Is the geometry of the plants the same, j
- 7. Is the type of fuel in all plants the same, ,
l
- 8. Are the fuel loading pattems in the plants similar (i.e., low leakage, etc.),
- 9. What is the projected 32 effective full power year surface fluence of each plant, l
- 10. What vessel inlet temperatures do the plants operate at,
- 11. What are the differences in the capsule lead factors of the plants, 4
- 12. Can the criteria for credibility in 10 CFR Part 50.61 be met for each plant?
April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves
. - - -. . . -. . . . .. -.~ -. ....-.- _ _. ......-.- -. _ .- - . - - -_
A-2 i EVALUATION: i 5 1. What weld wire heat number, flux and flux lot numbers were used to fabricate the welds? \ The sumeillance program weld metal for each unit was fabricated with the following weld wire and flux: Byron 1: The weld metal is type Linde MnMoNi, heat number 442002, with a Linde 80 type flux, not number 8873. This is the same heat number used in the limiting beltline weld (seam WF 336). l } Byron 2: The weld metal is type Linde MnMoNi, heat number 442002, with a Linde 80 i type flux, lot number 8064. This is the same heat number used in the limiting l
- beltline weld (seam WF-447).
t l The Byron Units 1 and 2 surveillance program weld metals were fabricated with the same i heat of weld wire and the same type of flux. Therefore, this information supports the l j integration of the suNeillance program test results for these welds. l i 2. What vendor fabricated the welds and in what time frame ? Byron 1: B&W fabricated the welds in the mid.1970's I Byron 2: B&W fabricated the welds in the mid.1970's The Byron Units 1 and 2 surveillance program weld metals were fabricated in the same time frame and by the same vendor. Therefore, this information supports the integration of the surveillance program test results for these welds.
- 3. What heat treatment did each weld receive?
The surveillance program weld metals received the following post-weld stress relief heat treatments: Byron 1: 1125 25'F for 12 hours and 16 minutes; fumace-cooled Byron 2: 1150
- 50*F for 13.5 hours; fumace-cooled The post-weld stress relief heat treatment given to the Byron 1 and 2 surveillance program welds was slightly different. However, based on engineering judgement, the slight differences in temperature and time should not cause a significant difference in the material toughness properties.
- 4. Is theinitialRTa of the nelds the same or relatively close?
Byron 1: -30 Y Byron 2: 10 Y Based on the data specific to the Byron 1 and Byron 2 vessel beltline welds (WF-336 and WF-447, respectrisly), the initial RT, y of the welds differ. However, the surveillance matenals have performed similarty, and it is shift data that is used in the integration of April 1997 Byron Unit 1 Hestup and Cooldown Limit Curves
. .- . - . . - ~ - _ - . - . . - - . ~ - - . - . ~ . _ . . .- - _. - - - - .-.
7 A3 {. data. As can be seen in Table 4 (page 10 of this report), the measured shifts in RT, are l relatively the same. For example, the shift for the first capsules from Byron 1 and Byron 2 ~ is O'F. For the second capsules removed from Byron Units 1 and 2, the measured shifts are equal to 30*F and 35'F, respectively. These results are very close. Therefore, this l information supports the integration of the surveillance program test results for these ', welds. i S. Is the initial ypper shelf energy of the surveillance welds the same or relatively close? Byron 1: 74 ft-lb i Byron 2: 67 ft-lb i
' The initial upper shelf energy values for the surveillance weld materials in the Byron i surveillance programs are very similar. Therefore, this information supports the integration
! of the surveillance program test results for these welds. i 6. Is the geometry of the plants the same? ! Byron Units 1 and 2 have a reactor vessel inner diameter of 173 inches, a reactor vessel beltline thickness of 8.5 inches (excluding the cladding). Both have a power rating of l 3411 MWt and are Westinghouse 4-loop NSSS plants. Both vessels have neutron pads a and the surveillance capsules are located at inc same azimuthal angles.
- 7. Is the fuel design in aIIplants the same? ~
Byron 1 & 2 use 17X17 rod array fuel assemblies with the same fuel design, thus l
- producing similar radiation effects at the surveillance capsules.
4
- 8. Are the fuelloading patterns in the plants similar (i.e. Iow leakage, etc.)?
Byron 1 & '2 use a low leakage loading pattom.
- 9. What is the prcrected 32 effectrve fuH power year surface fluence of each plant?
Based on the information provided below, the projected vessel surface fluence values
- (E>1.0 MeV) at 32 EFPY for Byron Unit 1 are essentially the same as Byron Unit 2. ;
if 29 39 49 f O' 1.947x10" 2.159x10" 1.705x10" 1.939x10" 1.290x10" Byron Unit 2 29 3F 49
! c' if 1.979x10" 2.192x10" 1.772x10" 2.026x10" J 1.353x10" l
i I 4 April 1997
; Byron Unit 1 Heatup and Cooldown Limit Curves
A-4 I
- 10. What are the vesselinlet temperatures?
Byron 1: 558.4*F Byron 2: 558.4*F
- 11. What are the differences in the capsule lead factors of the plants? l Based on the information provide in Table 1, the lead factors of the surveillance capsules i j in Byron Unit 1 are essentially the same as Byron Unit 2.
TABLE A-1 Surveillance Capsule Lead Factors for Byron Units 1 & 2 s Byron Unit 1 Byron Unit 2 Capsule Location Lead Factor Caosule Location Lead Factor i U 58.5 3.85 U 58.5 3.96 X 238.5 3.79 W 121.5 3.89 V 61.0 '3.59 V 61.0 3.64 Y 241.0 3.59 Y 241.0 3.64 W 121.5 3.79 X 238.5 3.89 Z 301.5 3.79 Z 301.5 3.89 Based on the projected vessel surface fluence and lead factor values for Byron 1 and 2, the Byron 1 and 2 surveillance capsules will have approximately the same flux rates and irradiation temperatures. This supports the use of the weld results from both programs to evaluate the reactor vesselintegrity of both units.
- 12. Can the criteria for credibility in 10 CFR Part 50.61 be met for each plant?
Credibility will be evaluated for 1) all the surveillance capsule data (base metal & weld metal) for Byron Unit 1, and 2) weld metal (only) for Byron Unit 2. The cred:bility determination will use the Byron Unit 2 weld metal data for the Byron 1 heatup/cooldown pressure-temperature limit curves. Therefore, it must be determined to be credible. Criterion 1: The materials in the surveIIIsnce capsules must be those which are controlling materials with regard to radiation embrittlement. The following is a list of the beltline materials contained in the Byron Units 1 and 2 surveillance programs: April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves
_ . _ . _ _ _ _ _ _ _ _ .._ _ _ _ ._. .. _ _ _ ___ __.__.m._._
- l. I
- A5 i Byron Unit 1
- Intermediate F. hell forging 5P-5933 J Circumferential weld seam WF-336, heat number 442002, with a Linde 80
, type flux, not number 8873. (This is the same heat number useo in the limiting bettline weld.) j } Byron Unit 2: Intermediate shell forging 49D329/49C297-1-1 Circumferential weld seam WF 447, heat number 442002, with a Linde 80 l . type flux, lot number 8064 (This is the same heat number used in the limiting beltline weld.) } I 1 i Based on the calculated RTm values presented in WCAP-13881 (Byron 1 PTS) and the information provided in the Byron Unit 2 material selection documents, these materials are j judged to be the most controlling with regard to radiation embrittlement for each unit. l Therefore, Criteria #1 is met for both units. h Criterion 2: ScatterIn the plots of Charpy energy vstaus temperature for the Irradiated and unitradiated conditions must be small enough to permit l the determination of the 30 N-lb temperature unambiguously. Plots of Charpy energy versus temperature for the unirradiated condition are presented in WCAP-9517, ' Commonwealth Edison Co. Byron Station Unit 1 Reactor Vessel Radiation Surveillance Program," dated July 1979 and WCAP-10398, " Commonwealth Edison Co. Byron Station Unit 2 Reactor Vessel Radiation Surveillance Program," dated December 1983. Plots of Charpy energy versus temperature for the irradiated conditions are presented in the WCAP reports for Capsules U & X (Unit 1) and U & W (Unit 2). Based on engineering judgement, the scatter in the data presented in these reports is small enough to determine the 30 ft-lb temperature and the upper sheff energy of the Byron Units 1
& 2 surveillance weld metals unambiguously. Therefore, the Byron Units 1 & 2 surveillance materials most this criteria.
Criterion 3: Where there are two or more sete of surveillance date kom one reactor, the scatter of ART, values must be ines than 26*F 16r welds and 17'F 16r bene motel. Even of the venge In the capsule Ruences is large (two or more ordere of magnitude), the ocetter may not exceed twice those values. i i The least squares method, as
- scribed in Regulatory Positen 2.1, will be utilized in l determining a best-fit line for this data to determine if this criteria is met.
l April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves
A-6 4 TABLE A-2* Byron Units 1 & 2 Surveillance Capsule Data Calculation of Best Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 Material Capsule f") FF*) Measured FF x FF' 2 (x) ART,3 ART,,or (x ) (y) (xy) Byron Unit 1 U 3.72x10 O.727 0 0 0.529 Inter. Shell Forging SP-5933 (Axial) X 1.39x10 1.091 30 32.73 1.190 Byron Unit 1 U 3.72x10 O.727 0 0 0.529 inter. Shell Forging 5P-5933 X 1.39x10 1.091 30 32.73 1.190 (Tangential) I ",,, 3.636 60 65.46 3.44 Byron Unit 1 U 3.72x10 O.727 0 0.00 0.529 Weld Metal
- X 1.39x10 1.091 35 38.185 1.190 Byron Unit 2 U 3.996x10 O.746 0 0.00 0.557 Weld Metal !
W 1.211x10 1.053 30 31.600 1.110 I" , 3.617 65 69.785 3.386 NOTES: (a) f = Fluence (10 n/cm ,a E > 1.0 MeV) (b) FF = Fluence Factor = f*'"*88 (c) Values of Iand ART, , for Byron 1 were taken from WCAP-14044 and WCAP 13880, respectively. The Byron Unit 2 values were taken from Table 3 of WCAP 14063. Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
_._ . . _- . ~ _ _ _ _ _ _ _ . . _ _ _ . . _ _ _ _ . . - _ _ _ . _ . _ _ . _ _ _ . _ _ . _ _ _ . _ 7 A-7 ; 1
\
Per the 27.* Edition of the CRC Standard Mathematical Tables (page 497), for a straight line ; fit by the method of least squares, the values b,and b, are obtained by solving the normal 1 l equations: n b,+ b, Ex, = Iy, l and - b, Ex,+ b, Ex,8 = Exy, . i' l These equations can be re-written as follows: ? n n y j E.i,=an+bEx, i.i I and n n n [ x,y, = a{ x, + b[ x,* i.i i1 i-1 l I Byron 1 & 2 Wold Metal: i Based on the data provided in Table A 2 the equations become: l
- 1. 65.0 = 4a + 3.617b or a = 16.25 - 0.9043b and
- 2. 69.785 = 3.617a + 3.386b Thus, by substituting Eq.1 into Eq. 2, b = 95.71. Now, enter b (= 95.71) into Eq.1 and a = -
70.30. Therefore, the equation of the straight line which provides the best fit in the sense of least squares is: , Y' = 95.71 (X) - 70.30 The error in predicting a value Y corresponding to a given X value is: e = Y - Y' Byron 1 Base Metal: Based on the data provided in Table A-2 the equations become:
- 1. 80.0 = 4a + 3.636b or a = 15.0 - 0.909b and
- 2. 65.46 = 3.636a + 3.44b Thus, by substitutmg Eq.1 into Eq. 2, b = 80.89. Now, enter b (= 80.89) into Eq.1 and a = -
58.53. Therefore, the equation of the straight line which provides the best fit in the sense of least squares is: Y' = 80.89 (X) - 58.53 The error in predicting a value Y corresponding to a given X value is: e = Y - Y' l l April 1997
. Byron Unit 1 Heatup and Cooldown Limit Curves :
)
A8 TABLE A-3 Best Fit Evaluation for Byron 1 & 2 Surveillance Materials Base Material ARTu, Best Fit Scatter of FF (30 ft-lb) (*F) ARTun (*F) ARTun ('F) unummmmmmmmeammmmmmmmmmmmm-mummmmmmmmmmmmu Byron 1 & 2 0.727 0 -0.72 -0.00 We!d Metal . 1.091' 35 -0.00 35.00 0.746 0 -0.00 0.00 1.053 30 -0.00 30.00 4 Byron Unit 1 Inter. Shell 0.727 0 0.28 -0.28 Forging SP-5933 (Axial) 1.091 30 29.72 0.28 Byron Unit 1 inter. Shsil 0.727 0 0.28 -0.28 l Forging SP-5933 (Tangential) 1.091 30 29.72 0.28 i Weld Metal: The scatter of ARTum vaiues about a best-fit line drawn, as described in Regulatory 3 Position 2.1, should be less than 28'F for weld metal. As chown above, the error is within 28'F of the best-fit line. Therefore, this criteria is met for the Byron Units 1 & 2 surveillance weld material. 4 l Base Material: { The scatter of ARTum values about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 17'F for base metal. As shown above. the error is within 17'F of the best-fit line. Therefore, this criteria is met for the Byron Unit 1 surveillance base I metal. ! See the following scatter plots for the Byron Und 1 base material and the Byron 1 and 2 weld metal. i Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
A-9 i 1 l Byron 1 Base i.Jaterial I 30 ,e 25 - ,. e Byron 1 SP-5933 (Axial)- I
/ Measured w 20 - / 3 Byron 1 SP-5933 (Axial)- Best-Fit g 7' g 15 / - . AByron 1 SP-5933 (Tangential)-
a
/ Measured
.I o.
3
/ ,x Byron 1 SP-5933 (Tangentel)-
Best-Fit 5-l
"' I 0
0 0.5 1 1.5 l Fluence (1919, E>1.0MeV) l i i i i
'l l
' I Byron 1 & 2 Wold Metal 35 ,&
- w. /
g' i 25 -
/
/ I l
- 20 - ! i e Byron 1 & 2 Wold Metal- l k
s 15 -
/ ; Measured
'nByron 1 & 2 weid Metal Best-i
/' l l Fit j!
f io . ,/ 5- / 0 I
/
w
. 0 0.5 1 1 5
-5 ' !
. Fluenos (1919, E>1.00 tov) i i
1 i Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
^ A-10 4 4 i Criterion 4: The irradiation temperature of the Charpy specimens in the capsule
' must equal the vessel wall temperature at the cladding / base metal f interface within +/- 25'F.
i
- The Byron Unit 1 & 2 surveillance capsules are located in the reactor between the neutron i pads and the vessel wall and are positioned opposite the center of the core. The test i capsules are in baskets attached to the neutron pad. The location of the specimens with j respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the j specimens experience equivalent operating conditions and will not differ by more than 25'F.
l Additionally, since the vessel inlet temperatures are the same, the irradiation temperatures will ) [ be the same. l 1 l Criterion 5: The surveillance dets for the correlation mar;itor materialin the capsule, if present, must fall within the scatter band of the data base for the material. } Byron Units 1 & 2 did not incorporate correlation monitor material in their surveillance program. Therefore, Criterion 5 is not applicable. RESULTS & CONCLUSIONS: Based on the evaluation performed above, it has been determined that there is sufficient data to support integrating the Byron Unit 1 weld metal surveillance data with Byron Unit 2 weld metal surveillance data. , 1 i I i Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
A-11 EFFECT OF WELD METAL INTEGRATION ON BYRON P-T LIMIT CURVES: Previous Previous New New Results Plant 1/4T ART 3/4T ART 1/4T ART 3/4T ART Byron 1 66.37*) 57.15** 70'* 60'" -- Curves at 8 EFPY FDRT/SRPLO-
- 009(94) 33.2 92.6 75.8 Current curves / PTS Byron 2 43.5 evaluation are NOT Curves at conservative. Using 16 EFPY weld metal integration
' will be more restrictive.
WCAP 14063 Byron Unit 2 curves to l be regenerated and l documented in WCAP 14881 l NOTES, (a) Even after weld metalintegration, still forging-limited. Wold metal integration has no effect. ! (b) Calculated at 8 EFPY. (c) Calculated at 12 EFPY. The new ART values for Byron Unit 2 are significantly larger. A reasonable applicability date cannot be determined. New curves are to be generated for Byron Unit 2. The results will be I documented in WCAP-14881, " Byron Unit 2 Heatup and Cooldown Curves for Normal Operation'. a 4 1 April 1997 Byron Unit 1 Heatup and Cookiown Limit Curves
A-12 EFFECT OF WELD METAL INTEGRATION ON BYRON PTS CALCULATIONS: i The weld metal integration CF values were calculated in Section 4 of this report. Specifically, the following weid metal CF values were used to determine the RTns values: l RG Position 1 CF RG Position 2 CF Byron Units 1 and 2 82.0'F 61.8'F The vessel material data used in the latest PTS evaluation reports""Iwas used in this ! evaluation. (No new material property values were calculated.) However, for the Byron Units 1 and 2 RTn, calculations at 48 EFPY, new fluence values were interpolated to 48 EFPY. l I The vessel surface fluence results reported in Section 6.0 of the latest Byron Unit 1 R and Byron Unit 2 f"1 surveillance capsule analysis reports were used. TABLE A-4 RTn, Values for Byron Unit 1 CF f'" FF*' RT,ew M ARTm RTn, Material ('F) ('F) (*F) (*F) (*F) 32 EFPY interrnediate Shell Forging 23.8 2.159 1209 40 28.8 28.8 97.6 SP-5933 Using surv. capsule data'* 19.1 2.159 1209 40 17 23.1 80.1 Lower Shell Forging 5P-5951 26 2,159 1209 10 11.4 31.4 72.8 Weld Metal WF 336 82.0 2.159 1209 30 56 99.1 125.1 Using surv. capsule data'* 61.8 2.159 1209 -30 28 74.7 72.7 48 EFPY intermediate Shell Forging 23.8 3238 1.309 40 31 2 31 2 102.4 SP-5933 Using surv. capsule data'* 19.1 3.238 1.309 40 17 25.0 82.0 Lower Shell Forging SP-5951 26 3.238 1.309 10 34.0 34.0 78.0 82.0 3238 1.309 -30 56 107.3 133.3 Wold Metal WF 336 Using sury capsule data'* 61.8 3.238 1.309 30 28 80.9 78.9 I l MUi n (a) 2.159 x 10" n/cm' (E>1.0 MeV) for 32 EFPY from Byron 1 PTS report (WCAP 13881). The following ! calculation to obtain the 48 EFPY fluence value: l 2.159x10" + f2.159x10" 3.807x10")*(48 32 EFPY) = 3.238x10" n/cm' 32 5.64 EFPY l (b) FF (Fluence factor) = f""**** i (c) Calculated using a CF based on survemance capsule data per Regulatory Guide 1.99, Revision 2. Position 2. April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves
A 13 l TABLE A-5 l RTrrs VALUES FOR BYRON UNIT 2 l l CF (*F) f'* FFS) RT,orm M('F) A RTprs RTris Material (*F) (*F) ('F) ] 32 EFPY Lower Shell Forging 32.2 2.192 1.213 -20 34.0 39.1 53.1 MK 24-3 Using surv. capsule data'" 19.8 2.192 1.213 -20 17 24.0 21.0 Inter. Shell Forging 20.0 2.192 1.213 -20 24.3 24.3 28.6 MK 24-2 l Cire. Weld Metal WF447 82.0 2.192 1.213 10 56 99.5 165.5 Using surv. capsule data'" 61.8 2.192 1.213 10 28 75.0 113.0 48 EFPY Lower Shell Forging 32.2 3.288 1.312 -20 34.0 42.2 56.2 MK 24-3 Using surv. capsule data'" 19.8 3288 1.312 -20 17 26.0 23.0 Inter. Shell Forging 20.0 3.288 1.312 -20 262 262 32.4 - l MK 24 2 Cire. Weld Metal WF447 82.0 3288 1.312 10 56 107.6 173.6 Using sury. capsule data'" 61.8 3.288 1.312 10 28 81.1 119.1
- NOTES, (a) 2.192 x 10* rVcm' (E>1.0 MeV) for 32 EFPY from Byron 2 PTS report (WCAP 14054). The following calculation to obtain the 48 EFPY fluence value:
2.192x10 + (2.192x10 3.174x10P(48 32 EFPY) = 3.288x10 tverrf 32 - 4.634 EFPY (b) FF (Fluence factor) = f*8* * *'N' (c) Calculated using a CF based n survemance capsule data per Regulatory Guide 1.99, Revision 2. Position 2. Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
_ _ .- - . . . . .. _ - _ _ . . _ . . . _ - . - . . . ~ . . .. . _ . . - _ . . .. . -. l i B-0 i . 4 APPENDIX B 4 l ! WELD METAL INTEGRATION FOR BRAIDWOOD UNITS 1 AND 2 4 2 l 1 i l ) i l l l l t i J l Byror. Unit 1 Heatup and Cooldown Limit Curves April 1997
B-1 4 INTRODUCTION: Westinghouse performed an evaluation to determine if the weld wire data of the Braidwood Units 1 and 2 surveillance programs can be integrated. The evaluation was based on the following criteria:
- 1. What weld wire heat number, flux, and flux lot were used to fabricate the surveillance program weld metal of.each unit,
- 2. What vendor fabricated the welds and in what time frame, ,
- 3. What heat treatment did each surveillance program weld receive,
- 4. Is the inMial RT,m of the welds the same or relatively close,
- 5. Is the initial upper shelf energy of the welds the same or relatively close,
- 6. It the geometry of tne plants the same, i
, 7. Is the type of fuel in all plants the same,
- 8. Are the fuel loading pattoms in the plants similar (i.e., low leakage, etc.),
- 9. What is the projected 32 effective full power year surface fluence of each plant,
- 10. What vessel inlet temperatures do the plants operate at,
- 11. What are the differences in the capsule lead factors of the plants,
- 12. Can the criteria for credibility in 10 CFR Part 50.61 be met for each plant?
1 i Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
. 1 B-2 J
]
" o i
EVALUATION: j i i l 1. What weld wire heat number, flux and flux lot numbers were used to fabricate the welds? Braidwood 1: The weld metal is classification EF2N Low Cu, MnMoNi Heat number ! 442011, with a Linde grade 80 type flux, lot number 8061. This is the ! same heat number used in the limiting beltline weld (seam WF-562). I Braidwood 2: The weld metal is classification EF2N Low Cu, MnMoNi Heat number 44201 with a Linde grade 80 type flux, lot number 8061. This is the same heat number used in the limiting beltline weld (seam WF-562). j
.The Braidwood Units 1 and 2 surveillance program weld metals were fabricated with the
- same heat of weld wire and the same type of 9ux. Therefore, this information supports the integration of the surveillance program test results for these welds. ;
{ l
- 2. What vendor fabricated the nolds and in what time frame 7 l Braidwood 1: B&W fabricated the wolds in the late 1970's l Braidwood 2: B&W fabricated the welds in the late 1970's l
The welds for Braidwood 1 and 2 were fabric 'd in the same time frame and by the I ome vendor. Therefore, this information supports the integration of the surveillance ; nrogram test results for these welds. - !
- 3. What heat treatment did each surveillance program weld receive?
Braidwood 1: 1100 - 1150*F for 12% hours; fumace cooled. Braidwood 2: 1150 50*F for 12% hours; fumace cooled. The post-weld stress relief heat treatment given to the Braidwood 1 and 2 surveillance program welds was slightly different. However, based on engineering judgement, the slight differences in temperature and time should not cause a significant difference in ths l material toughness properties.
- 4. Is the initialRTuor of the welds the same or relatively close?
Braidwood 1: 40 Y Braidwood 2: 40 Y The Braidwood Units 1 and 2 initial RT,, values are identical. Therefore, this information supports the integration of the surveillance program test results for these welds. syron unit 1 Hestup and cooisown Lirnit curves Ardi1997
j -
- B-3
- 5. Is the initial upper shelf energy of the surveillance welds the same or relatively close?
Braidwood 1: 70 ft lb Braidwood 2: 71 ft-Ib l 4
. The initial upper shelf energy values for the surveillance weld materials in the Braidwood
! surveillance programs are very similar. Therefore, this information supports the integration of the surveillance program test results for these welds. 1 l , j 6.' is the geometry of the plants the same? j All four plants have a reactor vessel inner diameter off 173 inches, a reactor vessel l beltline thickr.e::s of 8.5 inches (excluding the cladding), and a NSSS 4-loop power rating of 3411 MWT. In addition, all four plants have neutron pads and the surveillance capsules are located at the same azimuthal angles. ( 7. Is the fuel design in sIIplants the same?
- Braidwood 1 & 2 use 17X17 rod array fuel r.stamblies with the same fuel design, thus l producing similar radiation effects at the curveillance capsules.
- 8. Are the fuelloading pattems in the plants similar (i.e. Iow leakage, etc.)?
l Braidwood 1 & 2 use is low leakage loading pattom. i k 9. What is the projected 32 eNective fullpoweryear surface Ruence of each plant? l ! Based on the information provided below, the projected vessel surface fluence (E>1.0 ) MeV) values at 32 EFPY for Braidwood Unit 1 are essentially the sama as Braidwood ) i Unit 2. I Braidwood Unit 1 l O' 19 2F SF _ 4F
- 1.321x10 1.984x10 2.239x10 1.86Ex10 2.162x10
1 Braidwood Unit 2 0' 15* 2F 3F 4F 1.299x10 1.924x10 2.199x10 1.861 x10 2.174x10
- 10. What noselidet temperatures do the plants operate?
Braidwood 1: 558.4*F Braidwood 2: 558.4*F
- 11. What are the clWerences in the capsule land factors of the plants?
Based on the information prende in Table B 1, the lead factors of the surveillance capsules in Braihood Unit 1 are essentially the same as Braidwood Unit 2. syron uri 1 Heatup ard Conidown Limit Curves April 1997
~
B-4 d TABLE B-1 Surveillance Capsule Lead Factors for Braidwood Units 1 & 2 Braidwood Unit 1 Braidwood Unit 2 Capsule Location Lead Factor Capsule Location Lead Factor U 58.5$ 4.03 U 58.5 4.00 l X 238.5 4.03 X 238.5 4.02 W 121.5 4.03 W 121.5 4.02 l Z 301.5 4.03 Z 301.5 4.02 V 61.0 3.73 V 61.0 3.70 Y 241.0 3.73 Y 241.0 3.70
- Based on the projected vessel surface fluence and lead factor values for Braidwood 1 & 2, the i Braidwood 1 & 2 surveillance capsules will have approximately the same flux rates and 1 irradiation temperatures. This supports the use of the surveillance weld data in both programs to evaluate the reactor vessel integrity of the Braidwood units.
- 12. Can the criteria for credibility in 10 CFR Part 50.61 be met for each plant?
t Credibility will be evaluated for the Braidwood Units 1 and 2 weld metal (only) to show that Braidwood 1 & 2 can share weld metal data and determine an integrated weld metal chemistry factor. j Criterion 1: The materials in the surveillance capsulea must be those which are ' controlling materials with regard to radiation embrittlement. The following is a list of the beltline materials contained in the Braidwood Units 1 and 2 surveillance programs: Braidwood Unit 1: Lower shell forging 49D867/49C813-1-1 Circumferential weld seam WF-562, heat number 442011, with a Linde grade 80 type flux, lot number 8061. (This is the same heat number used in the limiting beltline weld.) Braidwood Unit 2: Lower shell forging 50D102/50C97-1-1 Circumferential weld seam WF-562, heat number 442011, with a Linde grade 80 type flux, lot number 6061. (This is the same heat number used in the limiting beltline wc ; Based on the calculated RTm values presented in 5 ' .1242 (Braidwood 1 PTS) and the information provided in the Braidwood Unit 2 material selection documents, these materials Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
i B5 J
~
i are judged to be the most controlling with regard to radiation embrittlement for each unit. i Therefore, Criteria #1 is met for both Braidwood units. 4 Criterion 2: Scatter in the plots of Charpy energy Versus temperature for the l Irradiated and unirradiated conditions must be small enough to permit i the determination of the 30 ft-Ik temperature unambiguously. , i Plots of Charpy energy versus temperature for the unirradiated condition are presented in l ! WCAP-9807, " Commonwealth Edison Company Braidwood Station Unit No.1 Reactor Vessel ! Radiation Surveillance Program," dated February 1981 and WCAP-11188, " Commonwealth l Edison Company Braidwood Station Unit No. 2 Reactor Vessel Radiation Surveliiance i Program," dated December 1986. Plots of Charpy energy versus temperature for the : irradiated conditions are presented in the WCAP reports for Capsules U & X for both ustts. } i I j Based on engineering judgement, the scatter in the data presented in these reporte is small ; j enough to determine the 30 ft-lb temperature and the upper shelf energy of the Braidwood j Units 1 & 2 surveillance weld metals unambiguously. Therefore, the Braidwood Units 1 & 2 surveillance materials meet this criteria. i Criterion 3: Where there are two or more sets of surveillance data from one reactor, l 1 the scatter of ARTm values must be has than 28*F for welds and 17'F . for knee metal. Even if the range in the capauk Ruences is large (two , t or more orders of magnitude), the scatter may not exceed twice those l values. l l The least squares method, as described in Regulatory Position 2.1, will be utilized in , ! determining a best-fit line for this data to determine if this criteria is met. - i . l l l 1 i i i ) I i l Byron Unit 1 Hestup and Cooldown Umit Curves April 1997 j i
r B-6 4 j TABLE B-2* Braidwood Units 1 & 2 Surveillance Capsule Dsta Calculation of Best Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 Material Capsule f FF*' Measured FF x FF' (x) ART , ART. (x') (y) (xy) Braidwood U 0.3814x10 O.733 10 7.333 0.538 Unit 1 i Weld Metal X 1.144x10" 1.038 25 25.95 1.077 Braidwood U 0.3933x10 O.741 0 0.00 0.550 , Unit 2 X 1.126x10 1.033 20 20.66 1.067 Weld Metal I", 3.545 55 53.943 3.232 I i Chemistry Factor) = 53.943 + 3.232 = 16.7 NOTES: (a) f = Fluence (10 n/cm , E > 1.0 MeV) (b) FF = Fluence Factor = f"""" men (c) Values of f, FF, and ART, values were taken from Table 2 of WCAP-14213 (Braidwood Unit 1 P. T Limg) and WCAP- 30 (Braidwood Unit 2 P T Limits). (d) CF = MFF'RT ) + 'FF') Per the 27 Edition of the CRC Standard Mathematical Tables (page 497), for a straight line fit by the method of least squares, the values b, and b, are obtained by solving the nonTial equations n b,+ bi Ex, = Ey, and b, Ex, + b, Ex,8 = Exy, These equations can be re-written as fo;;ows: 1 ' n n [ y, = an + b[ x,
! i. i.i 4
- and i n n n 1
[ xy, = a[ x, + b[ x,
..i ..i ..i Byron Unit 1 Heatup and Cooldown Limit Curves April 1997
B-7 b Braidwood 1 & 2 Weld Metal: Based on the data provided in Table B-2, the equations become: I 1.) 55.0 = 4a + 3.545b or a = 13.75 - 0.8863b and 2.) 53.943 = 3.545a + 3.232b Thus, by substituting Eq.1 into Eq. 2, b = 57.69. Now, enter b (= 57.69) into Eq.1 and a = - 4 37.38. Therefore, the equation of the straight line which provides the best fit in the sense of j least squnres is: Y' = 57.69 (X) - 37.38 i The error in predicting a value Y corresponding to a given X value is: e = Y - Y' TABLE B-3 Best Fit Evaluation for Byron 1 & 2 and Braiowood 1 & 2 Weld Mete! Base Material ART , Best Fit Scatter of FF (30 ft-lb) ('F) ARTom (*F) A RT , (*F) arrns-umummmuummmmmuums-mummummmmmmmmmmu Braidwood 1 & 2 0.733 10 -0.00 10.00
* * "' 1.038 25 -0.00 25.00 0.741 0 -0.00 0.00 1.033 20 -0.00 20.00
( Weld Metal: 1 The scatter of ARTum values about a best fit line drawn as described in Regulatory Position 2.1, should be less than 28'F for weld metal. As shown above, the error is within 28'F of the best-fit line. Therefore, this criteria is met for the Braidwood Units 1 & 2 surveillance weld material. See the following plot of ARTum versus fluence. l Byron Unit 1 Heatup and Cooldown Umit Curves April 1997
j B-8 i . i l Braidwood 1 & 2 Wold Metal l )- j 25 0 1 20 - + l . I * ' e Breedwood i & 2 Weld Metal-
- g 15 - Measured E 3 Braidwood 1 & 2 Weld Metal-10 , , Best Fst 1
5- ) 0 0 4 0 0.2 OA 0.6 0.8 1 1.2 Fluence (1019, E>1.0MeV) i ) Criterion 4: The irradiation temperature of the Charpy specimens in the capsule I should onetch the vessel wall temperature at the cladding / base metsi l Intertece within sf- 25'F. l l t The Braidwood Unit 1 & 2 surveillance capsules are lccated in the reactor betwcen the j neutron pads and the vessel wall and are positioned opposite the center of '.he core. The test j capsules are in baskets attached to the neutron pad. The location of the rpecimens with ( respect to the reactor vessel beltline provides assurance that the reactor sessel wall and the i specimens experience equivalent operating conditions and will not differ by more than 25'F. , j Additionally, since the vessel inlet temperatures are the same, the irradiation temperatures will l be the same. I Criterion 5: ' The surveillance dets llor the correlation monitor material in the capeuie shoukt fall within the scatter band of the data base for that meterial. Braidwood Units 1 & 2 cid not incorporate correlation monitor material in their surveillance program. Therefore, Critorion 5 is not applicable. RESULTS & CONCLUSIONS: Based on the evaluation performed above it has been determined that there is sufficient data to support integrating the Eraidwood Unit 1 weld metal surveillance data with Braidwood Unit 2 wold metal surveillance data. Byron Unit 1 Hostup and Cooldown Limit Curves April 1997
i~ B-9 i . f 4 TABLE B-4 i Calculation of Average Cu and Ni Weight Percent Values for the Braidwood l Weld Material (Using Braidwood 1 & 2 Chemistry Test Results) i i > Best-Estimate ! Reference Qg M l j B&W Weld Qualification BAW 2261 0.028 0.63 , B&W Wold Qualification 0.03 0.65 B&W Wold Qualification 0.04 0.6 / ! l 0.032 0.671 -> 0.03 0.67 Sury. CF = 41 Braidwood 1 Surv. Data See Below
-See Below 0.033 0.706 -> 0.03 0.71 Sury. CF = 41
- Braidwood 2 Surv. Data Best-Estimate Chemistry; 0.033 0.666 -> 0.03 0.67 Best Est. CF = 41 Standard Deviation
- 0.005 0.029 Braidwood 1 & 2 Ratio = 1.0 l l
Surveillance Chemistry Results: Braidwood UnN 1 Braidwood UnN 2 Reference pdg g Refemnoe Q! M WCAP 9807 0.04* 0.67* WCAP-11188 0.040 0.64 WCAP 12685 0.035 0.006 WCAP 14228 0.033 0.724 0.033 0.666 0.034 0.711 0.034 0.723 0.033 0.714 1 0.035 0.700 0.038 0.780 0.034 0.728 0.035 0.737 l 0.035 0.099 0.033 0.728 1 0.035 0.751 0.032 0.752 ( 0.031 0.683 0.G32 0.743 O.032 0.673 0.031 0.730 0.029 0.068 0.032 0.711 0.029 0.686 0.032 0.728 . 0.034 0.616 0.031 0.703 0.033 0.651 0.032 0.687 0.033 0.006 0.033 0.703 0.031 0.666 0.033 0.005 , WCAP-14241 0.031 0 r55 WCAP 12845 0.032 0.704 O.029 0.647 0.034 0.754 0.028 0.838 0.032 0.008 0.031 0.065 0.026 0.623 0.031 0.860 0.028 0.835 0.032 0.061 0.031 0.679 0.033 0.067 0.029 0.644 0.028 0.648 0.uS2 0.099 0.027 0.644 0.034 0.765 0.034 0.008 0.031 0.673 0.033 0.056 0.034 0.724 0.036 0.868 0.035 0.747 0.036 0.671 0.033 0.711 M M 0.031 0.888 i Averspo 0.032 0.871 0.035 0.750 9.81 M Average 0.033 0.708
- Not used in Awarage ceiculeton, reported for completeness The same value appears in the material test suposts and the survemenos program esport.
Byron Unit 1 Hestup and C0oldown Limit Curves April 1997
B 10 4 EFFECT OF WELD METAL INTEGRATION ON BRAIDWOOD P-T LIMIT CURVES: Plant Previous Previous New New Result 1/4T AP- 3/4T ART 1/4T ART 3/4T ART 1 Braidwood 1 76.6 65.4 69.7 60.6 Current curves / PTS Curves at evaluation are 16 EFPY conservative. WCAP-14243 New Applicability Date: 27.9 EFPY Braidwood 2 62.6 55.7 69.5 60.4 Current curves / PTS Curves at evaluation are NOT conservative. Using l 1S EFPY ' weld metal integration WCAP-14230 will be more restrictive. ! New Applicability Date: 7.4 EFPY After the Braidwood Units 1 and 2 surveillance weld metal is integrated, the following calculations show the new applicability dates of the heatup/cooldown pressure-temperature limit curves. BRAIDWOOD UNIT 1: Weld Metal calculations based on a 1/4T ART = 76.6'F: (The following data is from Braidwood Unit 1 heatup/cooldown curve report, WCAP-14243) Per Regulatory Guide (RG) 1.99, Revision 2: ART = 1 + M + (CF
- FF)
Using the " Previous' ART v lues and initial RT , this equation was used to back-calculate the fluence factor (FF) and the vessel surface fluence value. This fluence value was then used to determine a new applicability date (in terms of EFPY) for the current pressure-temperature limit curves. For Braidwood Units 1 and 2, the mca.i. term from the above equation was calculated as (CF'FF) in the latest heatup/cooldown curve WCAP report. The following text explains this methodology from Regulatory Guica 1.99, Revision 2. April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves
\
s E B-11 l i
" 8 Ti ne Margin term is calculated as, M = 2 (o,' + o3 ). The standard deviation for the initial RTum margin term (o) is O'F when the initial RTuor is a measured value (as is the case for l the Byron units). Additionally, the temi o, need not exceed 0.5 time:, the mean value of
- A RTum, i l
} Therefore, when the ART , value is multiplied by.0.5 and plugged into the above squation,. ! [ the effect is 2 * (ART , /2), which is the ART ,(or CF
- FF).
I i' ART = 1 + (CF
- FF) + (CF
- FF) !
76.6*F = 40*F + (16.7
- 1/4T FF)*F + (16.7
- 1/4T FF)*F ==> 1/4T FF = 1.0958 1.0958 = 1/4T f*" * *d "' "' 8 ==> 1/4T f = 1.4124 x 10 n/cm8 1.4124 x 10 = f *Me '" * ** ==> f = 2.352 x 10 n/cm' This fluence value will occur after 32 EFPY, per Table 6-15 of WCAP-14241. The following calculation will determine the applicability date in terms of EFPY.
Fluence at X EFPY = Fluence at 32 EFPY + (X - 32 EFPY)
- Fluence /EFPY 2.352 x 10 = 2.239 x 10 + (X - 32 EFPY) * (2.239 x 10- 1.120 x 10'S 32 - 16 EFPY X = 33.6 EFPY I
1 1 l Bymn Unit 1 Heatup and Cooldown Limit Curves April 1967
. . . .__ . _ _ _ _ _ _ _ _ _ ~ _ _ _ _ . - _ . . _ __ _-. . . . _ . . .. ..__
!' B-12 f Wald Metal calculations based on a 3/4T ART = 65.4*F: (The following data is from Braidwood Unit 1 heatup/cooidown curve report, WCAP-14243)
- ART = 1 + M + (CF
- FF) l 65.4*F = 40*F + (16.7
- 3/4T FF)*F + (16.7
- 3/4T FF)*F ==> 3/4T FF = 0.76047 i
0.76047 = 3/4T f" *' * * ==> 3/4T f = 0.4221 x 10 n/cm' 1 0.4221 x 10 = f
- e* "* ** ==> f = 1.9493 x 10 n/cm' i
.' This fluence value will occur between 16 and 32 EFPY, per Table 6-15 of WCAP-14241. The j following calculation will determino the applicability date in terms of EFPY. Fluence at X EFPY = Fluence at 16 EFPY + (X - 16 EFPY)
- Fluence /EFPY 5
} 1.9493 x 10 = 1.120 x 10 + (X - 16 EFPY)
- P M9 x 10- 1.120 x 10'S l 32 - 16 EFPY
! X = 27.9 EFPY 1 I Therefore, after the weld metal integration for Braidwood Units 1 and 2 is implemented, the i Braidwood Unit 1 heatup/cooldown curves presented in WCAP-14243 will be applicable to l 27.9 EFPY. i h
. Byron Unit 1 Hostup and Cooldown UmR Curves April 1997 i
4_-m.th%-a#- A 5 -.a w- 4 4A. 4 Ae- J .J+J(h.-+.a 44 4 ,4.Sud..-F---41 .A.+-* 4 .* re- S4Sa+.-d 4.16 -heae.d.d .2AJ#,, ad.A4.-.bbu,m. ..e-s*Aw.m 4--.ad-endI M->- J G44 m'a + Ab;-daea a d :.4E 4-me. #&eb+,q 4 L. E -13 i I BRAIDWOOD UNIT 2 !
- WakLMetal calculations based on a 1/4T ART = 62.6*F
(The following data is from Braidwood Unit 2 heatup/ coo!down curve report, WCAP-14230.) ART = 1 + M + (CF
- FF) l
- 62.6*F = 40*F + (16.7
- 1/4T FF)*F + (16.7
- 1/4T FF)*F ==> 1/4T FF = 0.6766 O.6766 = 1/4T f*" 'd * "8 ==- 1/4T f = 3.075 x 10" n/cm' l- 3.075 x 10" = f
- e(** * ** ") ==> f = 5.120 x 10" n/cm' l This fluence value will occur between 4.215 and 16 EFPY, per Table 615 of WCAP-14228. j i The following calculation will determine the applicability date in terms of EFPY.
i i, Fluence at X EFPY = Fluence at 4.215 EFPY + (X - 4.215 EFPY)
- Fluence /EFPY 5.120 x 10" = 2.896 x 10" + (X - 4.215 EFPY) * (1.100 x 10"- 2.896 x 10'$
16 - 4.215 EFPY X = 7.4 EFPY Wetd Metal calm!ations based on a 3/4T ART = 55.7'F: ART = 1 + M + (CF
- FF) 55.7'F = 40*F + (16.7
- 3/4T FF)*F + (16.7
- 3/4T FF)*F ==> 3/4T FF = 0.47006 0.47006 = 3/4T f*" 'd ""' ==> 3/4T f = 0.1292 x 10" n/cm' O.1292 x 10" = f
- e(** * " * ** ==> f = 5.966 x 10" n/cm' This fluence valua will occur between 4.215 and 16 EFPY, per Table 6-15 of WCAP-14228.
The following calculation will determine the applicability date in terms of EFPY. Fluence at X EFPY = Fluaw.;e at 4.215 EFPY + (X - 4.215 EFPY)
- Fluence /EFPY 5.966 x 10" = 2.896 x 10" + (X - 4.215 EFPY) * (1.100 x 10"- 2.696 x 10'S 16 - 4.215 EFPY X = 8.7 EFPY After the weld metal integration for Braidwood Units 1 and 2 is implemented, the Braidwood Unit 2 heatup/cooldown curves presented in WCAP-14230 will be apphcable to 7.4 EFPY.
Byron Unit 1 Hestup and Cooldown t.imR Curves April 1997
B-14 EFFECT OF WELD METAL INTEGRATION ON BRAIDWOOD PTS CALCULATIONS: The weld metal integration CF values were calculated in Section 4 of this report. Specifically, the following weld metal CF values were used to determine the RTers values: RG Position 1 CF RG Position 2 CF Braidwood Units 1 and 2 41.0'F 16.7'F The vessel material data used in the latest PTS evaluation reports"'#5 was used in this avaluation. (No new material property values were calculated.) TABLE B-4 RTn, Values for Braidwood Unit 1 CF f FFi') RT,er, M ARTprs RT,rs Material (*F) ('F) (*F) (*F) (*F) 32 EFPY inter. Shell Forging 31.0 2.239 1.218 -30 34 37.77 41.8 Lower Shell Forging 26.0 2.239 1.218 -20 31.68 31.68 43.4 using S/C data *' 18.8 2.239 1.218 -a 17 22.90 19.8 Weld Metal WF 562 41.0 2.239 1.218 40 49.95 49.95 139.9 using S/C data *3 16.7 2.239 1.218 40 20.34 20.34 80.7 i
! 48 EFPY inter. Shell Forging 31.0 3.358 1.317 -30 34 40.83 44.8 i
Lower Shell Forging 26.0 3.358 1.317 -20 34 34.25 48.3 1 using S/C data") 18.8 3.358 1.317 -20 17 24.76 21.8 41.0 3.358 1.317 40 54.00 54.00 148.0 Weld Metal WF 562 using S/C data *8 16.7 3.358 1.317 40 21.99 21.99 84.0 NOTES. (a) FF (Fluence factor) = f**N (b) Calculated using a CF based on survoitance capsule data per Regulatory Guide 1.99. Revision 2. Position 2. April 1997 Byron Unit i Heatup and Cooldown Limit Curns
B-15 TABLE B 5 RTers Values for Braidwood Unit 2 CF ('F) f FF) RTwow, M(*F) A RT,rs RT,73 Material (*F) ('F) ('F) 32 EFPY Upper Shell Forging 20.0 2.199 1.214 -30 24.28 24.28 18.6 Lower Shell Forging 37.0 2.199 1.214 -30 34 44.92 48.9 [ using S/C data') 13.3 2.199 1.214 30 16.15 16.15 2.3 Weld Metal WF 562 41.0 2.199 1.214 ' 40 49.77 49.77 139.5 using S/C data *> 16.7 2.199 1.214 40 20.27 20.27 80.5 48 EFPY Upper Shell Forging 20.0 3.298 1.313 30 26.26 26.26 22.5 l Lower Shell Forging 37.0 3.298 1.313 -30 34 48.58 52.6 i using S/C data") 13.3 3.298 1.313 -30 17 17.46 4.5
- Weld Metal WF 562 41.0 3.298 1.313 40 53.83 53.83 147.7 using S/C data *' 16.7 3.298 1.313 40 21.93 21.93 83.9
.N.,Q1gf.;
(a) FF (Fluence factor) = f** **** (b) Calculated using a CF based n surveillance capsule data per Regulatory Guide 1.99, Revision 2. Position 2. l i dpril1997 Byron Unit 1 Heatup and Couldown Umit Curves
- C-0
- APPENDIX C i
i 8 BYRON /BRAIDWOOD FLUENCE METHODOLOGY JUSTIFICATION 4 AND TIME-DEPENDENT CAPSULE FLUENCE VALUES a l i i i i i j Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 l J
C-1 1 - Fluence Methodoloav Justification The fast neutron exposure methodology documented in WCAP-14040 NP A, " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"is consistent with the requirements of Draft Regulatory Guide DG-1053, " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence' and makes use of neutron transport cross-sections derived from the ENDF/B-VI data base. The exposure evaluations documented in WCAPs 13880,14064,14241, and 14228 for j the Byron Units 1 & 2 and Braidwood Units 1 & 2 pressure vessels were completed prior to the islease of the ENDF/B-VI based Light Water Reactor neutron transport cross section f
- library. Consequently the neutron transpcrt calculations performed as an integral part of these
) evaluations were based on the currently available ENDF/B-IV based cross section library. In all respects other than the ENDF/B-VI vs ENDF/B-IV cross-section issue, the methodology l applied to the Byron Units 1 & 2 and Braidwood Units 1 & 2 fluence evaluations was identical i to the approved methodology described in WCAP-14040-NP-A. ) It is planned that neutron fluence evaluations for the Byron Units 1 & 2 and Braidwood Units 1
& 2, pressure vessels will be updated to incorporate the use of ENDF/B VI cross-section libraries at the time of the next scheduled surveillance capsule withdrawal for each of the units. Based on recent experience in updating vessel fluence evaluations to the ENDF/B-Vi methodology for reactors of similar design to Byron and Braidwood it is anticipated that best estimate neutron fluence values will be impacted by less than 7% compared to those over those previously reported. Likewise, application of the ENDF/B-VI methodology to the re-evaluation of neutron dosimetry from previously withdrawn surveillance capsules will change reported capsule exposures by an amount less than the uncertainties quoted with the prior dosimetry analyses.
In addition to the methodology upgrade diameaad in the preceding paragraph, the fluence updates for Byron Units 1 & 2 and Braidwood Units 1 & 2 will also include an evaluation of low leakage fuel management instituted at all four units. A qualitative examination of the loading pattems used at Byron Units 1 & 2 and Braidwood Units 1 & 2 indicates that accounting for the flux reduchon brought about by the incorporation of low leakage fuel management will compensate for increases in projected fluence that may be introduced by the methods changes. The not effect of methods upg ades and low leakage fuel management on Bron Unit 1 Heatup and Cooldown Limit Curves April 1997
)
. _ _ _. . _ . . . _ _ - . _ . ~ _ _ _ _ . _ _ . _ . _ . - _ . . . . _ ._.__.- . _ _
i Co2 l l projected vessel fluence is, therefore, anticipated to be very small and may result in an overall ! reduction in fluence relative to that reported in WCAPs 13880,14064,14241, and 14228.
- Based on the relatively small changes that are anticipated from updating the neutron fluence evaluations from those reported in WCAPs 13880,14064,14241, and 14228 to the approved methodology described in WCAP-14040-NP-A, including the impact of low leakage fuel f
j management, coupled with the low sensitivity to irradiation damage exhibited by the materials comprising the Byron Units 1 & 2 and Braidwood Units 1 & 2 reactor pressure vessels, the l use of the previously documented fluence values is justified until the update to the ENDF/B-Vi based methodology is completed for each unit. A 2 - Time n nandent Survaillance Caosule Fluences i Based on the documentation provided in WCAPs 13880,14064,14241, and 14228, it is noted ! that the last surveillance capsule withdrawal for Byron Units 1 & 2 and Braidwood Units 1 & 2 was at 5.64,4.63,4.23, and 4.21 effective full power years, respectively. Projection of fluence levels at the surveillance capsule locations for times beyond those withdrawal dates are j needed in order to establish appropnate withdrawal schedules for the remaining capsules l ! comprising the Reactor Vessel Surveillance Program for each of the units. These Best '
- Estimate projections are provided in Tables C-1 through C-4 for Byron Units 1 & 2 and 1
i Braidwood Units 1 & 2, respectively. These projections are based on the assumption that the f best estimate neutron flux averaged over the total irradiation time. for each unit would remain ' ! applicable for the remainder of plant lifetime. 1 i i i April 1997 Byron Urut 1 Hostup and Cooldown Limit Curves
C-3 i . TABLE C-1 BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS ) i AT SURVEILLANCE CAPSULE LOCATIONS - BYRON UNIT 1 l l Irradiation Fluence [n/cm'] Lead Factor i Time
- M 31.5 Caos 29.0 Cads 31.5 Caos 29.0 Caos j 5.64 1.443e+19 1.365e+19 3.79 3.58 I 8.00 2.046e+19 1.935e+19 3.79 3.58 10.00 2.558e+19 2.419e+19 3.79 3.58 12.00 3.070e+19 2.902e+19 3.79 3.58
- 14.00 3.581e+19 3.386e+19 3.79 3.58 16.00 4.093e+19 3.870e+19 3.79 3.58 l 18.00 4.604e+19 4.353e+19 3.79 3.58
~
20.00 5.116e+19 4.837e+19 3.79 3.58 l 22.00 5.628e+19 5.321e+19 3.79 3.58 1 24.00 6.139e+19 5.804e+19 3.79 3.58 l 26.00 6.651e+19 6.288e+19 3.79 3.58 1 1 28.00 7.162e+19 6.772e+19 3.79 3.58 ! 30.00 7.674e+19 7.256e+19 3.79 3.58 32.00 8.186e+19 7.739e+19 3.79 3.58 1 .. h Byrdn Unit 1 Heatup and Cooldown Limit Cunss April 1997
M C-4 TABLE C2 BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS AT SURVEILLANCE CAPSULE LOCATIONS - BYRON UNIT 2 8 Irradiation Fluence (n/cm ] Lead Factor i Time IEFPY1 31.5 Caos 29.0 Caos 31.5 Caos 29.0 Caos 4.63 1.235e+19 1.154e+19 3.89 3.64 6.00 1.598e+19 1.494e+19 3.89 3.64 8.00 2.131e+19 1.992e+19 3.89 3.64 10.00 2.664e+19 2.491e+19 3.89 3.64 12.00 3.197e+19 2.989e+19 3.89 3.64 14.00 3.730e+19 3.487e+19 3.89 3.64 I l 16.00 4.262e+19 3.985e+19 3.89 3.64 ! 18.00 4.795e+19 4.483e+19 3.89 3.64 20.00 5.328e+19 4.981e+19 3.89 3.64 22.00 5.861e+19 5.479e+19 3.89 3.64 24.00 6.394e+19 5.077e+19 3.89 . 3.64 26.00 6.927e+19 6.475e+19 3.89 3.64 28.00 7.459e+19 6.973e+19 3.89 3.64 30.00 7.992e+19 7.472e+19 3.89 3.64 32.00 8.525e+19 7.970e+19 3.89 3.E 4 l l i Byron Unit 1 He=tip and Cooldown Limit Curves April 1997
. _ . _. - - .. m.. _ . _ _ . _ _ . _ _ . _ . _ _ . . _ . _ . . _ . .... _ . _ _ . . _ _ _ . _ . _ . . _
4 C5 4 l TABLE C-3 1 l BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS AT SURVEILLANCE CAPSULE LOCATIONS - BRAIDWOOD UNIT 1 l l l Irradiation Fluence [n/cm'] Lead Factor Tm l' IFJfI 31.5 Caos 29.0 Caos 31.5 Caos 29,0 Caos 4.23 1.193e+19 1.105e+19 4.02 3.73 l 6.00 1.690e+19 1.565e+19 4.02 3.73 j 8.00 2.254e+19 2.087e+19 - 4.02 3.73 10.00 2.8170+19 2.609e+19 4.02 3.73 12.00 3.380e+19 3.130e+19 4.02 3.73 l 14.00 3.944e+19 3.652e+19 4.02' 3.73 16.00 4.507e+19 4.174e+19 4.02 3.73 18.00 5.070e+19 4.696e+19 4.02 3.73 I 20.00 5.634e+19 5.217e+19 4.02 3.73 22.00 6.1970+19 G.739e+19 4.02 3.73 24.00 6.761e+19 6.261e+19 4.02 3.73 26.00 7.324e+19 6.783e+19 4.02 3.73 4 28.00 7.887e+19 7.304e+19 4.02 3.73 30.00 8.451e+19 7.826e+19 4.02 3.73 32.00 9.014e+19 8.348e+19 4.02 3.73 Byron Unit 1 Hestup and Cooldown Umit Cunes April 1997
C-6 . TABLE C-4
- BEST ESTIMATE F ST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS l AT SURVEILLANCE CAPSULE LOCATIONS - BRAIDWOOD UNIT 2 i
i i i } Irradiation Fluence [n/cm'] Lead Factor Time
- [EFPY1 31.5 Caos 29.0 Caos 31.5 Caos 290 Caos '
1 l 4.21 1.163e+19 1.072e+19 4.02 3.70 a' 6.00 1.656e+19 1.526e+19 4.02 3.70 j 8.00 2.208e+19 2.034e+19 4.02 3.70 ! 4.02 10.00 2.760e+19 2.543e+19 3.70 [ 12.00 3.312e+19 3.051e+19 4.02 3.70 1 14.00 3.864e+19 3.560e+19 4.02 3.70
- 16.00 4.416e+19 4.068e+19 4.02 3.70 18.00 4.968e+19 4.577e+19 4.02 3.70 20.00 5.520e+19 5.085e+19 4.02 3.70 22.00 6.072e+19 5.594e+19 4.02 3.70 l
I 24.00 6.625e+19 6.102e+19 4.02 3.70
-26.00 7.177e+19 6.611e+19 4.02 3.70 28.00 7.729e+19 7.119e+19 4.02 3.70 30.00 8281e+19 7.628e+19 4.02 3.70 32.00 8.833e+19 8.136e+19 4.02 3.70 o
Bymn Unit 1 Hestup and Cooldown Limit Curves April 1997
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