ML20202F839
| ML20202F839 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 11/30/1997 |
| From: | Shaun Anderson, Christopher Boyd, Laubham T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20202F824 | List: |
| References | |
| WCAP-14824, WCAP-14824-R02, WCAP-14824-R2, NUDOCS 9712090239 | |
| Download: ML20202F839 (80) | |
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Westinghouse Non-Proprietary Class 3 k + +++++++
D ,
BYRON UNIT 1 ,
HEATUP AND COOLDOWN LIMIT CURVES FOR >
NORMAL OPERATION AND SURVEILLANCE WELD METAL .
INTEGRATION FOR BYRON & BRAIDWOOD Westinghouse Energy Systems 54 fr$A# $$$$0 $5b$
WESTINGHOUSE NON-PROPRIETARY CLASS 3
\" CAP-14824, Revision 2
)
) Byron Unit 1 Heatup and Cooldown Limit Curves For Normal Operation and Surveillance Weld Metal Integration
)
for Byron and Braidwood T. J. Laubham
)
S. L. Anderson November 1997
)
Work Performed Under Shop Order CPEP-139
) Preparca by the Westinghouse Electric Corporation for the Commonwealth Edison Company
) Approved: .
C. H. Boyd, M4ageT Engineering & Materials Technology
)
WESTINGHOUSE ELECTRIC CORPORATION Nuclear Services Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
)
C 1997 Westinghouse Electric Corporation All Rights Reserved k
l
. . . . ._. o
. l
)
PREFACE
)
This report has been technically reviewed and verified by:
S. Abbott
[, Mk I
p
-?
)
)
I
)
)
. Byron Unit i Heatup and Oooldown Limit Curves November 1997
li TABLE OF CONTENTS LIST OF FIGURE . . . ... ... ... . . .. . . . .. . . . .. . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . iii !
LI ST O F TAB L E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. ... .. ... .. ... .. .. . .. .. . .. .. ... .. .. . . iv 1
i N T R O D U C Tl O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. . . . . . . . . . . . . . . . . .
2 FRACTURE TOUG HN ESS PROPERTIES............ .. .. ....... ..... ... ....... ..... ... ..... 2 3
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3 4
CALCULATION OF ADJUSTED REFERENCE TEMPERATURE. . ..... . .. ....... 6 5
HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES. .. . 16 6
REFERENCES......................................................................... 24 APPENDIX A - WELD METAL INTEGRATION FOR BYRON UNITS 1 AND 2.... . .......... A-0 APPENDIX B - WELD METAL INTEGRATION FOR BRAIDWOOD UNITS 1 AND 2.... .. B-0 APPENDIX C - BYRON /BRAIDWOOD FLUENCE METHODOLOGY JUSTIFICATION AND TIME-DEPENDENT CAPSULE FLUENCE VALUES... . . . . . . . . . C-0 Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
~
lii h
LIST OF FIGURES
)
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 i n stru me ntation E rro rs) . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . .18
)
1 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 i n strume nt ation E rro rs) . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
)
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 Errors; Margin of 74 psig for Pressure Difference Between
?
Pressure instrumentation and the Reactor Vessel Beltline Region)....................... 20 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)....................... 21
)
)
)
)
)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
y . 1 1 INTRODUCTION k 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 materialin the core region of the reactor vesselis determined by
. using the unirradiated reactor vessel material fracture toughness properties, estimating the j radiation-induced ARTuor, 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.
) RTuor increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTuor at any time period in the reactor's life, ARTuor due to the radiation exposure associated with that time period must be added to the unirradiated RTuor(IRTuor). The extent of the shift in RTuor is enhanced by certain chemical elements (such as copper and nickel) present y 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 Embrittlement of Reactor Vessel Materials'". Reguiatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTuor + 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 metalinterface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves, j
Y
)
)
)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
2
)
0 2 FRACTURE TOUGHNESS PROPERTIES
)
The fracture toughness properties of the ferritic materialin the reactor ecolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan:t1. The pre-irradiation fracture toughness properties of the Byron Unit 1 reactor vessel are presented in Table 3. Credible surveillance data is available for two capsules (Capsules U and X) for Byron
) Unit 1. The post-irradiation fracture toughness properties of the reactor vessel beltline material were obtained directly from the Byron Unit 1 Reactor Vessel Radiation Surveillance Program f*l.
This capsule data is used to calculate chemistry factors (See Table 4) in addition to those calculated per Regulatory Guide 1.99, Revision 2.
Additionally, per the request of the Commonwealth Edison Company, the surveillance weld data from the Byron Unit 1 and Byron Unit 2 surveillance programsH3 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 l resulting from the weld metalintegration for Byron Units 1 and 2 are presented in Section 4 of l this report. See Tables 1 through 4.
A complete technical justification for the Byron Units 1 and 2 weld metal integration is presented
) in Appendix A of tnis report.
l
! Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
. --..: as6. n.i _i-_
3 ?
3 CRIT 2RIA FOR ALLOWABLE PRESSURE-TEMPERATURE
) RELATIONSHIPS Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements"W specifies fracture toughness requirements for ferritic materia's 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 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 Plant Components"N,
) Vessels, contain the conservative methods of analysis.
The ASME approach for calculating the allowable limit curves for various heatup and coaldown rates specifies that the total stress intensity factor, K , for the combined thermal and pressure 3
stresses at any time during heatup or cooldown cannot be greater than the reference stress
) intensity factor, Ki., for the metal temperature at that time. K, i is obtained from the reference fracture toughness curve, defined in Appendix G of the ASME Code, Section Xim. The K, curve is given by the following equation:
)
Ku = 26.78 + 1.233
- ci m xr-a w nim (3) where,
) K,i = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTuor Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
)
C
- K i + Ka < Ku (2)
J where, K. = stress intensity factor caused by membrane (pressure) stress K,i = stress intensity factor caused by the thermal gradients K,i = function of temperature relative to the RTuov of the material C= 2.0 for Level A and Level B service limits C= 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical
) Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
4
- O At any time during the heatup or cooldown transient, the allowable value K is determined by the metal temperature at the tip of a postulated flaw at the il4T and 3/4T location, the appropriate value for RT,m, and the reference frac ture toughness cury.9 The thermal stresses 8 resulting from the temperature gradients through the vessehvali are calculated and then the corresponding (thermal) stress intensity f6ctors, K , for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated, g
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses ct the inside, which increase with 9 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.
The use of the composite curve in the cooldown analysis is necessary bect.use control of the O cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessellocation is at a higher temperature inan the fluid adjacent to the vesselinner diameter. This condition, of course, is not true for the steady state g situation. it follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher allowable value of N at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist sa that the increase in allowable value K exceeds Kn, the calculated allowable pressure during cooldown will be greater than the steady-state value. O 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 developed 9 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 intemal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the allowable value K. for the 1/4T crack during heatup is lower than the allowable value K for the 1/4T c ack during steady-state conditions at tne 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 Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 l
) .
- 5 allowable K values do not chet 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 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 steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure temperature
)
limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vesselinside 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 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 finallimit curves are produced by constructing a composite curve Maed on a point-by point comparison of the steady state and finite heatup rate data. At any give.n temperature, the allowable pressure is taken to be the lesser of the three valuen taken from the curves under consideration. The use of the composite curve is necessary to set 3 conservative heatup limitations because it is possible for conditions to exist wherein, over the F course of the heatup ramp, the controlling condition 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 regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTuor by at least 120*F for norma l operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is 621 psig for Byron Unit 1.
)
The limiting unirradiated RTuor 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.
)
)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 l
)
6 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE
)
From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each I materialin the beltline region is given by the following expression: l ART = InitialRTar + b RTnr + Margin (3)
)
Initial RTm7i s the reference temperatuie for the unirradiated material as defined in paragraph l
NB-2331 of Section 111 of the ASME Boiler and Pressure Vessel Code'). If measured values of initial RTer for the materialin 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 standard
)
deviation for the class.
ARTmy is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
k 6 RTar = CF
- l'"***"*' zgy
) To calculate ARTe1 at any depth (e.g., at 1/4T or 3/4T), the foilowing formula must first be used to attenuate the fluence at the specific depth.
f p m = f ,,, ,,* e/ * (5)
)
where x inches (vessel beltline thickness is 8.5 inchest "l) is the depth into the vessel wall measured from the vessel clad / base metalinterface. 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 shell forgings and circumferential weld at 12 EFPY is 8.10 x 10" 2
n/cm . This fluence value was calculated from the surveillance Capsule X analysis presented in WCAP-13880I'l.
)
Explanation for the Application of the Credibility Criteria and the Ratio Procedure in calculating RTris values in accordance with 10 CFR 50.61 (which incorporates Regulatory Guide 1.99 Revision 2 in total) and ART values for input to 10 CFR 50 Appendix G pressure-
) temperature limit curves, Commonwealth Edison (Comed) uses the methodology described in Regulatory Guide 1.99 Revision 2. When there are two or more sets of surveillance data available, which there is in this case, Regulatory Guide 1.99 Revisic.i 2 provides criteria for
)-
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
7 .
g evaluating the credibility of the surveillance data and subsequently a procedure for determining a best-fit chemistry factor which represents the actual behavior of the material, normalized to the vessel of interest. O in particular, the third credibility criteria from Reg. Guide 1.99 Rev. 2 states that the " scatter of ARTuor values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28'F for welds and 17'F for base metal." This evaluation of credibility ,
becomes a comparison between the actual measured surveillance data shifts, and a line drawn using Equation 2 from Reg. Guide 1.99 Rev. 2 with a Regulatory Position 2.1 least-squares-fit chemistry factor based on the actual surveillance data set. This provides an indication of whether the surveillance material, with its specific measured chemistry, is behaving as the Reg.
Guide 1.99 Rev. 2 Equation 2 Irradiation damage correlation would predict. 8 Thus, for welds if it is determined that the surveillance data is credible relative to Reg. Guide 1.99 Rev. 2, and "if there is clear evidence that the copper or nickel content of the surveillance weld differs from that of the vessel weld, i.e., differs from the average [ currently taken to be the
'best estimate' chemistry) for the weld wire heat number associated with the vessel weld and .he surveillance weld,"N then for subsequent RTers and ART calculations, the ratio procedure of l Regulatory Position 2.1 is used to normalize the observed behavior cf the surveillance material to the expected behavior of the vessel weld. The measured values of ARTuo7 obtained from surveillance data are adjusted by multiplying them by the ratio of the Regulatory Position 1.1 e (Table 1) chemistry factor for the vessel weld to that for the surveillance weld. The ratio-adjusted surveillance data is used to calculate a least-squares-fit chemistry factor appropriate to ths vessel weld.
l The chemistry factors (CF, 'F) obtained from the tables in Reg. Guide 1.99 Revision 2 using the O average values of copper and nickel content as calculated in Tables 1 and 2 are reported in Table 3. The chemistry factors were also calculated using the surveillance capsule data in Table 4.
9 The Ratio Procedure, as documented in Regulatory Guide 1.99 Revision 2 Position 2.1, was used to adjust the measured values of ARTuor 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.
O All materials in the beltline region of Byron Unit 1 reactor vessel were considered in determining the limiting material and the calculations to determine the ART values at 12 EFPY are shown in Table 5. The resulting ART values for all beltline region materials at the 1/4T and 3/4T locations are summarized in Table 6, where it can be seen that the limiting material is the Intermediate g
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) wi!!
be used in the generation of heatup and cooldown curves applicable to 12 EFPY.
Byron Unit 1 Heatup and Cooldown Limit Curves Y
November 1997 l
. 8 TABLE 1 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1 Base Materials Reference Intermediate Shell Forging SP- Lower Shell Forging 5933 SP-5951 Cu%
Ni% Cu% Ni%
Byron Unit 1 0.034 0.73 0.04 0.64 HU/CD Limit Curves 0.032 0.791 i
0.03 0.75 i
Letter Report 0.05 0.73 FDRTl SRPLO-09(94) 0.036 0.735 January 1994
- Average 0.0364 0.747 0.04 0.64 Standard Deviation 0.007 0.023 0 0 l
1 t
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
9
- C 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) #
Best-Estimate Reference Cu Ni B&W Weld Qualification BAW-2261 0.024 0.7 B&W Weld Qualification 0.031 0.46 B&W Weld Qualification 0.03 0.72 B&W Weld Qualification 0.068 0.48 B&W Weld Qualification 0.114 0.54 B&W Weld Qualification 0.148 0.6 B&WWeld Qualification 0.053 0.62 B&W Weld Qualification 0.059 0.62 B&W Weld Qualification g
Ref. 23 0.029 0.65 Byron 1 Surveillance Datr See Below 0.022 0.690 -> 0.02 0.69 Surv. CF = 27 Byron 2 Surveillance Data See Below 0.023 0.712 -> 0.02 0.71 Surv. CF = 27 Best-Estimate Chemistry
- 0.055 0.617 -> 0.05 0.62 Best Est. CF = 68 Standard Deviation: 0.042 0.091 Byron 1 & 2 Ratio = 2.5'4 g Surveillance Data Chemistry Results: Byron Unit 2 Byron Unit 1 Reference Cu Ni Reference Cu Ni WCAP-10398l4} 0.03 0.65 WCAP-9517(3) 0.026 0.71 WCAP-12431[22] o 024 0,740 WCAP-11651[21) 0.023 0.67 0.024 0.786 0.022 0.665 0.022 0.704 0 0.021 0.714 0.020 0.691 0.021 0.741 0.021 0.706 0.022 0.713 0.020 0.697 0.021 0.714 0.019 0.668 0.020 0.704 0.022 0.759 0.020 0.694 0.021 0.714 g 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 0.680 0.021 0.744 0.021 0.680 0.022 0.738 0.021 0.667 0.022 0.771 0.024 0.677 WCAP-14064l11] 0.024 0.705 9 0.022 0.697 0.023 0.706 0.021 0.634 0.023 0.698 WCAP-13880l9} 0.024 0.682 0.024 0.696 0.022 0.678 0.023 0.711 0.025 0.705 0.024 0.708 Avera9e 0.022 0.690 0.024 0.716 0.024 0.715 0 0.024 0.707 0.024 0.720 0.024 0.717 0.024 0.711 0.024 0.706 0.024 0.707 g 0.025 0.717 Average 0.023 0.712 Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 d l
)
10 TABLE 2 NOTES:
(a) The weld materialin 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 Byron Unit 2 survel. lance 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, lot number 8064.
) (c) Actual ratio is 2.5 (68.0 + 27.0 = 2.5), however, for conservatism a ratio of 3.0 wdl be used herein.
(d) The best estimate chemistry values was obtained using the " average of averages" approach.
TABLE 3
) Byron Unit 1 Reactor Vessel Material Properties Material Description Cu(%) Ni (%) Chemistry Initial Factor
- RTuor (*F)*
Closure Head Flange -
0.74 -
60N 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 0.05 0.62 68.0 -30 NOTES:
e (a) Chemistry Factors are calculated from Ce and Ni values per Regulatory Guide 1.99, Revision 2, Position 1.1
) (b) Mitial RTuor values are measured values.
(c) Closure head and vessel flange initia! RTuor values are used for considering flange requirementsW for the heatup/cooldown curves.
)
D
')
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 !
l
11 '
C TABLE 4 Calculation of Chemistry Factors Using Credible Byron Units 1 and 2 Surveillance Capsule Data #
Material Capsule Capsule FFIS Meas. FF* FF2 Fluence f ARTuor ARTuor e
Inter. Shell Forging SP-5933 U 3.72x10 O.727 0 0 0.529 (Tangential)
X 1.39x10 1.091 30 32.73 1.19 Inter. Shell rorging U 3.72x10 O.727 0 0 0.529 SP-5933 X 1.39x10 1.091 30 32.73 1.19 (Axial)
Sum: 65.46 3.44 9
Chemistry Factor 4* = 65.46 + 3.44 = 19.0*f Byron 1 Weld Metal U 3.72x10 O.727 0 0 0.00 0.529 WF-336*)
I X 1.39x10" 1.091 35 105(* 114.56 1.19 Byron 2 Weld Metal U 3.996x10 O.746 0 0 0.00 0.557 WF-447'"
4 W 1.211x10 1.053 30 90'* 94.77 1.110 Sum: 209.33 3.386 Chemistry Factor (* = 209.33 + 3.386 = 61.8'F g
NOTES:
(a) FF = Fluence Factor = f 28 0WO (b) Byron Unit 1 ARTuo, values were obtained from the surveillance 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-140440'l. 4 (c) Byron Unit 2 capsule fluence, FF, and ARTuorvalues were obtained from the surveillance Capsule W analysis (WCAP-14064i"l) using the ENDF/B-V scattering cross sections.
(d) Chemistry Factor = I(FF*ARTum) + E(FF2)
(e) Adjusted ARTuor per Ratio Procedure of RG1.99R2 Position 7 ' Ratio = 3.0 (See Table 2). Actual ratio is 2.5 (68.0 + 27.0 = 2.5), however, for conservatism a ratio of 3.0 was used in this case. As for 4 adjustments due to temperature difference between Units 1 and 2, see Appendix C page C-5 for explanation.
Byron Unit i Heatup and Cooldown Limit Curves d
November 1997 l
12 Explanation of Margin Terms used for Byror. Unit 1 When there are 'two or more credible surveillance data sets"N available for Byron Unit 1, Regulatory Guide 1.99 Rev. 2 (RG1.99R2) Position 2.1 states *To calculate the Margin in this case, use Equation 4; the values given there for cA may be cut in half". Equation 4 fron-RG1.99R2 is as follows: M = 2da: , g; . The values of oa are referred to as "28'F for
) welds and 17'F for base n etals."
Standard Deviation for Initial RTwor Margin Term, oi Since the initial RTwo, values are measured values in the case of Byron Unit 1, then o,is taken to be O'F s
) Standard Deviation for ARTuor Margin Term, ca Per RG1.99R2 Position 1.1, the values of og are referred to as *28'F for welds and 17'F for base metal, except that og need not exceed 0.50 times the mean value of ARTuor." The *mean value of ARTuor"is defined in RG1.99R2 by Equation 2 and defined herein by Equation 4. The
)
chemistry factor in RG1.99R2 Equation 2 is calculated from Tables 1 and 2 of RG1.99'C.
Per RG1.99R2 Position 2.1, when there is credible surveillance data, ca is taken to be the lesser of % ARTuo1 or 14*F (28'F/2) for welds, or 8.5'F (17'F/2) for base metal. tATuor again
) is defined herein by Equation 4, while utilizing a *Best-Fit Chemistry Factor" calculated in accordance with Position 2.1 of RG1.99R2 and shown herein on Table 4.
Summary of the Margin Term Since o,is taken to be zero when a heat specific measured value of u,9ial RTuor t.re available (as they are in this case), the total margin term, based on Equation 4 of RG1.99R2, will be as follows:
)
. Position 1.1: Lesser of ARTuo, or 56*F for Welds Lesser of ARTuor or 34'F for Base Metal
. Position 2.1: Lesser of ARTuor or 28'F for Welds
) -
Lesser of ARTuor or 17'F for Base Metal
)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 !
o
13
- C
\
The following is a sample calculation of the margin term for the weid metal at the % T location.
The results for this calculation as well as the results for the remaining reactor vessel beltline materials are documented in Table 5. Q Margin Tean for Weld Metal (1/4T Location):
- From Equation 8
where, CF = 68.0 (R.G. Position 1.1)
= 61.8 (R.G. Position 2.1; i.e. using Surv. Caps. Data)
FF = 0.799 (@ 12 EFPY and Fluence = 8.10 x 10" n/cm') C Therefore, ARTuor = 54.30 (R.G. Position 1.1)
= 49.40 (R.G. Position 2.1; i.e. using Sury. Caps. Data)
- From Equation 4 (of R.G.1.99 R2)
- Af = 2do)' + o.'
where, % ARTuor = 27.15 (R.G. Position 1.1)
= 24.70 (R.G. Position 2.1; i.e. using Surv. Caps. Data) C o, = 0'F (Initial RTuor is Measured) og = Lesser of (% ARTuoy ) or (28'F) c
= 27.15 (R.G. Position 1.1) og = Lesser of (M ARTuor ) or (14'F)
= 14.00 (R.G. Position 2.1; i.e. using Sury. Caps. Data)
C Therefore, 2 Af = 2l0 + 27.15' = 54.30 (R.G. Position 1.1) 2 Af = 240' + 14.0 = 28.00 (R.G. Position 2.1; i.e. using Sury.
Caps. Data) C C
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
14 TABLE S Calculation of Adjusted Reference Temperatures (ART) at 12 EFPY for all Byron Unit 1 Reactor Vessel Material (based on credible surveillance capsule data)
Reactor Ves:;el Belt!ine Matenal f @ 12" Region Location Identifica5cn Cu% Ni% CFM EFPY Lt f* %-t FF I ARTc" r, c. M ART *'
(x 10") %-t f Lt FF
.r
% T Calculation
.ntermediate SheE Forging SP-5933 0.0364 0.747 23 8 0.810 l 0.486 0.799 40 19.0 0 9.5 19.0 78 Intermediate SheH Forging 19.0 0 810 0 486 0.799 40 15 2 0 7.6 15.2 70 4 using S/C Data lower sheB Forging 5P-5951 0.04 0 64 26 2 0.810 0.486 0.799 10 20.8 0 10.4 20.8 52 Gath Weld Metal WF-336 0.06 0.61 68 0 0 810 0.486 0.799 -30 54.3 0 27.15 54.3 79 Girth Weld Metal 61.8 0810 0.486 0.799 -30 49 4 0 14.0 28 0 47
-+ using S/C Da'a l
% T Calculation
~~
Ir termediate SheB Forgog SP-5933 l0.0364 0.70 23 8 0.810 0.175 0.538 40 118 0 64 12.8 66 Intermediate Shed Forging 19.0 0.810 0.175 0.538 40 102 0 5.1 10.2 60 4 using S/C Data lower she5 Forging SP-5951 0.04 0 64 26 0 0.810 0.175 0.538 10 14 0 0 7.0 14 0 38 Girth Weld Metal WF-336 0.06 0 61 68.0 0.810 0.175 0.538 -30 36.6 0 18.3 36.6 43 Girih Weld Metal 61.8 0.810 0.175 0.538 -30 33.2 0 14.0 28.0 31
-+ using SIC Data l NOTES:
(1) The Byron Unit 1 reactor vessel war thickness is 8.5 inches at the bettEne region.
(b) Fluence, f,is based upon fsurf (1019 n/cm2. E>1.0 MeV) = 0.810 at 12 EFPY.
(c) ART = 1 + ART,cr + M (This value was rounded per ASTM E29, using the " Rounding Method".)
(d) ARTeor = CF
(;) The CF is integrated between the Byron 1 Weld (WF-336, heat 8 442002) and the Byron 2 Weld (v.F-447. Heat 8 442002)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
15 '
.C 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 Position 1(a))
9 using credible surveillance 70*) 60*
capsule data (RG Pealtion 2(a))
Lower Shell Forging SP 5951 52 38 (RG Position 1(a))
1 Circumferential Weld WF-336 79 43 (RG Position 1(a))
using credible surveillance 47 31 capsule data (RG Position 2(a)) C 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. C k
d l
l Byron Unit i Heatup and Cooldown Limit Curves November 1997
) 16 4
5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT
)
CURVES Pressure temperature limit curves for normal heatup and cooldown of the primary reactor l
coolant system have been calculated for the pressure and temperature in the reactor vessel )
beltline region using the methodsD21 discussed in Section 3 and 4 of this report. The 1989 edition
)
methodology is also presented in WCAP 14040 NP AD*l, dated January 1996.
Since indication of reactor vessel beltline pressure is not available on the plant, the pressure difference between the wide range pressure transmitter and the limiting beltline region must 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 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 for Byron Unit 1D51. Figures 3 and 4 do include this pressure difference of 74 psig.
j Figures 1 and 3 present the heatup curves without margins for instrumentation errors using a heatup rate of 100'F/hr applicable for the first 12 EFPY. Figures 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.
j The reactor must not be made critical until pressure temperature combinations are to the right of the criticality limit line shown in Figures 1 through 4. The straight-line portion of the criticality 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 governing equation for the hydrostatic test is defined in Appendix G to Section XI of the ASME Code as follows:
1.5Kw < Ku (6) where, Kn is the stress intensity factor covered by membrane (pressure) stress, Kp 26.78 + 1.233 e 8'""'*""* ,
T is the minimum permissible metal temperature, and RTa is the metal reference nil-ductility temperature The criticatify limit curve specifies pressure temperature limits for core operation to provide additional margin during actual power production as specified in Reference 5. The Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
17 .
O pressure temperature limits or core operation (except for low power physics tests) are that the l 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 h
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 i reactor vessel at 12 EFPY is 203'F at 2485 psig. The verticalline drawn from these points on the pressure temperature curve, intersecting g a curve 40'F higher than the pressure temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 1 through 4 define all of the above limits for ensuring prevention of nonductile failure for the Byron Unit i reactor vessel. The data pohts used for the heatup and cooldown pressure- C 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 requirements for the Byron Unit i 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 RT,a of the material stressed by the boltup. Therefore, since the most limiting initial RT,a value is 60'F (closure head flange), the reactor vessel can be bolted up at this temperature. g 4
e 4
4 Byron Unit 1 Heatup and Cooldown Limit Curves 9
November 1997
18 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP 5933 (using surv. capsule dat.)
LIMITING ART VALUES AT 12 EFPY: 1/4T, 70'F 3/4T, 60'F 2500 ' .'
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0 50 100 150 200 250 300 350 400 450 500 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 i Heatup and Cooldown Umit Curves November 1997
19 g MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP 5933 (using sury. capsule data) #
LIMITING ART VALUES AT 12 EFPY: 1/4T. 70*F 3/4T. 60'F 2500 I ' '
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FIGURE 2 Byron Unit i Reactor Coolant System Cooldown Limitations (Cooldown e Rates up to 100'F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors)
I 1
Byron Unit 1 Heatup and Cooldown Umit Curves November 1997
) 20 MATERIAL PROPERTY BASIS
) LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP 5933 (using surv. casule data)
LIMITING ART VALUES AT 12 EFPY: 1/4T, 70*F 3/4T, 60'F 2500 i, , , . ,, , .
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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 Errors; Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and the Reactor Vessel Beltline Region)
)
Byron Unit i Heatup and Cooldown Limit Curves November 1997
21 -
.d l
MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP 5933 (using surv. capsule data)
LI:JiT ING ART VALUES AT 12 EFPY: 1/4T, 70'F 3/4T, 60'F 2500 , ,
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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 Heatup and Cooldown Limit Curves d
November 1997 1
)
TABLE 7 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 psig per 10CFR50.
Cooldown Curves Heatup Curve Steady State 25F SOF 100F 100F Criticality, Limit Leak Test Limit
)
T P T P T P T P T P T P T P 60 621 60 595 60 554 60 470 60 621 203 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 90 621 90 607 90 621 203 646 95 621 95 621 95 621 95 621 95 621 203 639 100 621 100 621 100 621 100 621 100 621 203 634 105 621 105 621 105 621 105 621 105 621 203 632 110 621 110 621 110 621 110 621 110 621 203 633 115 621 115 621 115 621 115 621 115 621 203 637
) 120 621 120 621 120 621 120 621 120 621 203 642 125 621 125 621 125 621 125 621 125 821 203 651 130 621 130 621 130 621 130 621 130 621 203 661 135 621 135 621 135 621 135 621 135 621 203 674 140 621 140 621 140 621 140 621 140 621 203 689 145 621 145 621 145 621 145 621 203 707
) 150 621 150 621 150 621 203 727 155 621 155 621 203 749 160 621 160 621 203 774 165 621 165 621 205 801 170 621 170 621 210 831 175 621 175 621 215 8 64 180 621 180 621 220 900
) 180 1483 180 900 225 938 185 1559 185 938 230 980 190 1640 190 980 235 1026 195 1728 195 1026 240 1075 200 1821 200 1075 245 1128 205 1921 205 1128 250 1186 210 2029 210 1186 255 1247
) 215 2143 215 1247 260 1313 220 2266 220 1313 265 1385 225 2397 225 1385 270 1461 230 1461 275 1543 235 1543 280 1630 240 1630 285 1724
) 245 1724 200 1825 250 1825 295 1933 255 1933 300 2048 260 2048 305 2171 265 2171 310 2302 270 2302 315 2441 275 2441
)
(Configuration #9393315685880 for Cooldown, #2756858809292 for Heatup)
)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
23 '
.O TABLE 8 Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins for Instrumentation Errors d !
Includes 1) Vesset fiange requirements of 180*F and 621 psl0 per 10CFR$0. and 2) Pressure adjustment of 74 psig to account for pressure difference between the wide-range pressure transmitter and the limiting belthne region of th9 reactor vessel.
Cooldown Curves Heatup Curve C Steady State 25F SOF 100F 100F Criticahty, Lirnit Leak Test Limit T P T P T P T P T P T P T P 60 547 60 521 60 480 60 396 60 547 203 0 182 2000 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 75 80 547 547 75 80 547 547 75 531 80 547 75 80 457 480 75 80 547 547 203 203 0
597 q
85 547 85 547 85 547 85 505 85 547 203 583 90 547 90 547 90 547 90 5.$3 90 547 203 572 95 547 95 547 95 547 95 547 95 547 203 565 100 547 100 547 100 547 100 547 100 547 203 560 ;
105 547 105 547 105 547 105 547 105 547 203 558 110 547 110 547 110 547 110 547 110 547 203 559 b 115 547 115 547 115 547 115 547 115 547 203 563 120 547 120 547 120 547 120 547 120 547 203 568 125 547 125 547 125 547 ' 125 547 125 547 203 577 130 547 130 547 130 547 130 547 130 547 203 587 135 547 135 547 135 547 135 547 135 547 203 600 140 547 140 547 140 547 140 547 140 547 203 615 145 547 145 547 145 547 145 547 203 633 {
150 547 150 547 150 547 203 653 155 547 155 547 203 675 160 547 160 547 203 700 165 547 165 547 205 727 170 547 170 547 210 757 175 547 175 547 215 790 C 180 547 180 547 220 826 180 1409 180 826 225 854 185 1485 185 864 230 906 190 1560 190 906 235 952 195 1654 195 952 240 1001 200 47 200 1001 245 1054 205 210 1847 1955 205 210 1054 1112 250 255 1112 1173 hi 215 2069 215 1173 260 1239 220 2192 220 1239 265 1311 225 2323 225 1311 270 1387 230 1387 275 1469 235 1469 280- 1556 240 1556 285 1650
{
245 1650 290 1751 250 1751 295 1859 255 1859 300 1974 260 1974 305 2097 265 2097 310 2228 270 2228 315 2367 C (Configuration #9395568588093 for Cooldown, #9291115685880 for Hestup)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
24 6 REFERENCES
)
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 MTEB 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.
l 3 WCAP-9517, " Commonwealth Edison Co. Byron Station Unit i Reactor Vessel Radiation
) Surveillance Program", J. A. Davidson, July 1979.
4 WCAP-10398, " Commonwealth Edison Co. Byron Station Unit 2 Reactor Vessel Radiation Surveillance Program", L. R. Singer, December 1983.
I 5 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", Federal Register, Volume 60, No. 243, dated December 19,1995, l
j 6 1992 Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, Appendix G,
) " Vessels".
7 1989 ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix G,
- Fracture Toughness Criteria for Protection Against Failurs" 8 1989 Section ll1, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, " Material for Vessels".
9 WCAP 13880," Analysis of Capsule X from tne Commonwealth Edison Company Byron
) Unit 1 Reactor Vessel Radiation Survell!ance Program", P. A. Peter, et al., January 1994, 10 WCAP 14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation", E.
P. Lippincott, April 1994.
)
11 WCAP 14064, " Analysis of Capsule W from the Crammonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program", P. A. Peter, et al., November 1994.
12 WCAP-7924 A, " Basis for Heatup and Cooldown Limit Curves", W. S. Hazelton, et al.,
April 1975.
Byron Unit i Heatup and Cooldown Umit Curves November 1997
25 '
.C 13 WCAP 14040 NP A, Revision 2,
- Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup anc' Cooldown Limit Curves", J. D.
Andrachek, et al., January 1996. C l
14 Ba'acock & Wilcox drawing numbers 184557E, Rev. 2; 185266E, Rev. 2; 185297E, Rev.
2: 185328E Rev. 2: " Reactor V *,sel 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", 1 P. A. Peter, November 1994. C 17 WCAP-13881," Evaluation cf Pressurized Thermal Shock for Byron Unit 1", P. A. Peter, January 1994, 18 WCAP 14054, " Evaluation of Pressurtzed Thermal Shock for Byron Unit 2", P. A. Peter, August 1995.
1 19 WCAP 14242," Evaluation of Pressurized Thermal Shock for Braidwood Unit 1", P. A.
Peter, March 1995. g 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 C 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 Surveillance Program", E. Terek, et al., October 1989. C
- 23. Nuclear Design Information Transmittal, NDIT No. BYR97 346, Rev.0, " Additional data point for weld wire heat number 442002 for incorporation into Table 2 of WCAP 14824 Rev.1", Dated 9/9/97. C
- 24. NDIT No. BYR97-395, Rev. O,
- Byron Station Historical Teoid Data for Units 1 and 2",
B.J. Adams of Comed Byron, Dated 10 13-97.
- 25. NDIT No. BRW DIT 97-321, Rev. O, "Braidwood Station Historical Teold Data for Units 1 C and 2", M.A. Gorski of Comed Braidwood, Dated 10 22 97.
Dyron Unit i Heatup and Cooidown Umit Curves November 1997
26
- 26. WCAP 9807, " Commonwealth Edison Co. Braidwood Unit 1 Reactor Vessel Radiation
) Surveillance Program" S.E. Yanichko, et. al., February 1981.
- 27. WCAP 11188, " Commonwealth Edison Co. Braidwood Unit 2 Reacter Vessel Radiation Surveillance Program", L. R. Singer, December 1986.
)
l t
)
)
)
)
Byron Unit 1 Heatup and Cooldown Umit Curves November 1997
A-0
)
) APPENDIX A WELD METAL INTEGRATION FOR BYRON UNITS 1 AND 2 Y
l l
)
1
)
)
~
).
l p
?
)
) Byron Unit i Heatup and Cooldown Limit Curves November 1997
)
. . A1 INTRODUCTION:
Westinghouse performed an evaluation to determine if the weld wire data of the Byron 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, L !
i 3. What heat treatment did each surveillance program weld receive,
)
- 4. la the initial RT,1 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 the plants the same,
- 7. la 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?
)-
)
)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
A2
) .
EVALUATION:
- 1. What weld wire heat number, flux and flux lot numbers were used Ic fabricate the welds?
) The surveillance program weld metal for each unit was fabricated with the following weld wire and flux:
Byron 1: The weld metalis type Linde MnMoNi, heat number 442002, with a Linde 80 type flux, lot number 8873. This is the same heat number used in the limiting beltline
) wald (seam WF 336).
Byron 2: The weld metal is type Linde MnMoNi, heat number 442002, with a Linda 80 type flux, lot number 8064. This is the same heat nt.mber used in the limiting beltline weld (seam WF-447).
) The Byron Units 1 and 2 surveillance program weld metals were fabricated with the same heat of weld wirs and the same type of flux. Therefore, this information supports the integration of the surveillance program test losults for these welds.
- 2. What vendorfabricated the welds andin what time frame ?
) Byron 1: D&W fabricated the welds in the mid.1970's Byron 2: B&W fabricated the welds in the mid.1970's The Byron Units 1 and 2 surve31ance 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 recelve?
The surveillance program weld inetals received the following post weld stress relief heat
) treatments:
Byron 1: 1125
- 25'F for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 16 minutes; fumace-cooled Byrori 2: 1150150'F for 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; furnace-cooled
) The post weld stress relief heat treatment given to the Byron 1 and 2 surve ' lance program welds was slightly different. However, based on engineering judgement, the slight differences in temperature and time should not cause a significant different;e in the material toughness properties.
) 4. Is the initialRTuor fothe welds the same or relatively close?
Byron 1: -30 'F Byron 2: 10'F Based on the data specific to the Byron 1 and Byron 2 vessel beltline welds (WF 336 and
) WF-447, respectively, with the same weld wire heat and different flux lots), the initial RT,a of the welds differ. However, the surveillance materials have performed similarly under
) Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
- yww-,.e ,. - - , -
I 1
A3 . j
. 1 i
irradiation, and it is irradicFon shift data that is used in the integration of data. As can be j seen in Table 4 (page 11 of this report), the measured 6.iifts in RTm are relatively the i same. For example, the shift for the first capsules from Byron 1 and Byrun 2 is O'F. For the second capsules removed from Syron Units 1 and 2, the measured shifts are equal to 30'F and 35'F, respectively. These results are very close. Therefore, this information supports tho integration of the surveillance program test results for these welds.
- 5. Is the initial upper shelf energy of the surveillance welds the same or relatively close? e Byron 1: 74 ft lb Byron 2: 67 ft lb The initial upper shelf energy values for the surveillance '. veld materials in the Byron surveillance programs are very similar. Therefore, this information supports the integration
- of the surveillance program test results for these welds.
- 6. Is the geometry of the plants the same?
Byron Units 1 and 2 have a reactor vesselinner diameter of 173 inches, a reactor vessel beltline thickness of 8.5 inches (excluding the cladding). Both have a power rating of 3411 d
MWt and are Westinghouse 4 loop NSSS plants. Both vessels have neutron pads and the surveillance capsules are located at the same azimuthal angles.
- 1. Is the fuel design in allplants the same? C Byron 1 & 2 use 17X17 rod array fuel assemblies with the same fuel design, thus producing similar radiation effects at the surveillance capsules.
- 8. Are the fuelloading pattems in the plants similar(i.e. Iowleakage, etc.)?
Byron 1 & 2 use a low leakage loading pattern.
- 9. What is the projected 32 effecuve fullpower 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. g Byron Unit 1 0* 15' 25' 35' 45' 1.290x10" 1.947x10" 2.159x10" 1.705x10" 1.939x10" C
Byron Unit 2 0' 15' 25' 35' 45' 1.353x10" 1.979x10" 2.192x10" 1.772x10" 2.026x10" C
Byron Unit i Heatup and Cooldown Limit Curves November 1997
. _ A-4
) .
- 10. What are lho vesselinlet temperatures? (Per Reference 2A)
W D CYCLE # Byron Unit 1 Tcold ('F) Byron Unit 2 Teold ('F)
)
1 557 (Cap. U)* 557 (Cap. U)* ' ,
2 557 551 3 551 551
)
4 551 551 (Cap. W)* .,
5 551 (Cap. X)* 551 6 551 551
) 7 551 551 8 551 --
- See explanation in Append:x C, page C 5.
)
- 11. What are the differences in the capsule lead factors of the plants?
Based on the information provide in Table 1, the lead factors of the surveillance capsules in Byron Unit 1 are essentially the same as Byron Unit 2.
> TABLE A 1 Surveillance Capsule Lead Factors for Byron Units 1 & 2 Byron Unit 1 Byron Unit 2 Capsule Location
) Lead Factor Capsule Location Lead Factor U 58.5o 3.85 U 58.So 3.96 X 238.So 3.79 W 121.So 3.89
) V 61.0o 3.59 V 61.0* 3.64 Y 241.0* 3.59 Y 241.0* 3.64 W 121.So 3.79 X 238.5o 3.89
) Z 301.So 3.79 Z 301.So 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.
) Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 v
A5 *
,C i
- 12. Can the criteria for credibh..y in 10 CFR Parf 50.61 be met for each plant?
- Credibility wi'l be evaluated for 3,i thn si'r/elliance capsule data (base metal & weld metal) for Byron Units 1 and 2.
O 7
Criterion 1: The materials In the surveillance capsules must be those which are controlling materinIs with regard to radiation embrittlement.
O The following is a list c' the beltline materitis contained in the Eyron Units 1 and 2 surveillance programs:
Byron Unit 1: Intermediate shell forging SP 5933 C'.rcumferential weld seem WF 336, heat number 442002, with a Linde 80 type e flux, lot number 8873. (This is the saine heat number used in the limiting beltline weld.)
Byron Unit 2: Lower shell forging 49D330/49C2981 1 Circumfarential y; eld corm WF-447, heat number 442002, with a Linde 80 typo flux, lot number 8064. (This is the sar"* heat numbr used in the limiting beltline weld.)
Baced on the information provided in the material selection documents, WCAP 9517 (Byron 1, o See Ref 3) and WCAP-10398 (Byron 2, See Ref 4), thete materials are judged to be the most controlling with regard to radiation embrittlement for each unit. Therefore, Citeria #1 is met for both units.
O Criterion 2: Scatterin the plots of Charpy energy versus temperature for the kradicted ard unkradlated conditions must be small eno.)gh to permit the determination of the 30 ft Ib temperstma unnmbiguously.
Plots of Charpy energy versus temperature for the unirradiated condition c.e prusented 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 Vesso, Radiation Surveillance Program," dated December 1983. Plots of Charpy energy versus temperature for the irradiated coriditions are presented in the WCAP G reports for Capsules U & X (Unit 1) and U 8 W n Init 2).
Based on engineering judgement. th6 scatter in the data presented in these reports is small enough to determine the 30 ft lb temperature and the upper shelf energy of the Byron Units 1 &
2 surveillance weld metals unambiguously. Therefore, the Byron Units I & 2 surveillance materials meet this criteria.
l Byron Unit 1 Heatup and Cooldown Limit Cui;es November 1997 h
A6 Criterion 3: Where there are two or more sets of surveillance data from one reactor, the scatter of ARTuor values must be less than 28 F for welds and 17 f for base metal. Even if the range in the capsule fluences is large (two y more \
b orders of magnitude), the scatter may not exceed twice those values.
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. It should be noted here
) that the ratio procedure is not applied in this instance since only surveillance capsule data is being evaluated,
)
)
)
)
)
)
) November 1997 Byron Unit i Heatup and Cooldown Limit Curves
A7 .
.C TABLE A 2*
Byron Units 1 & 2 Surveillance Capsule Data Chemistry Factor for Best Fit Line 4
Material Capsule f' FFN Measured FF x FF8 ARTwot ARTwo, Byron Unit i U 3.72x10 O.727 0 0 0.529 Inter. Shell Forging 4/
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 X 1.39x10" 1.091 30 32.73 1.190 g l
l SP 5933 Sum: 65.46 3.44 l (Tangential) Chemistry Factor = 65.46 + 3.44 = 19,0'F Byron Unit 2 U 3.996 x 10 O.74S 0 0 0.557 Lower Shell Forging #
49D330-1/49C298-1 W 1.211 x 105 1.053 5 5.27 1.109 l (Axial)
Byron Unit 2 U 3.996 x 10" O.746 25 18.65 0.557 Lower Shell Forging w 3,233 x jois 1.053 40 42.12 1.109 49D3301/49C2981 Sum: 66.04 3.332 (Tangential) Chemistry Factor = 66.04 + 3.332 = 19.8'F Byron Unit 1 U 3.72x10 O.727 0 0.00 0.529 4
Weld Metal X 1.39x10 1.091 35'* 38.185 1.19 WF-336 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 0 WF-447 Sum: 69.785 3.386 Chemistry Factor = 69.785 + 3.386 = 20.6'F NOTES: 9 2
(a) f = Fluence (10" n/cm , E > 1.0 MeV)
(b) FF = Fluence Factor = f re.oi wn (c) Values of f ond ARTwo, for Cyron 1 were taken from WCAP 14044 and WCAP-13880, respectively.
The Byron Unit 2 values were taken from Table 3 of WCAP-14063.
(d) See Appendix C, page C-5, for explanatior, of temperature adjustment. g Byron Unit i Heatup and Cooldown Limit Curves November 1997
u A-8 TABLE A-3 l Best Fit Evaluation for Byron 1 & 2 Surveillance Materials Base Material CF .
FF ARTser Best Fit'* Scatter of < 17'F (Base Metals)
(30 ft-Ib) (*F) ART, , (*F) ART,cr (*F) < 28'F (Weld ".ietal)
Byron 1 20.6 0.727 0 1b.0 -15.0 Yes Weld Metal 20.6 1.091 35 22.5 12.5 Yes Byron 2 20.6 0.746 0 15.4 -15.4 Yes Weld Metal 20.6 1.053 30 21.7 8.3 Yes Byron Unit 1 Inter.
19.0 0.727 0 13.8 -13.8 Yes Shen Fagig SP-5933 (Axial) 19.0 1.091 30 20.7 9.3 Y-s Byron Unit 1 Inter.
19.0 0.727 0 13.8 -13.8 Yes l Shen Forging SP-5933 19.0 1.091 30 20.7 9.3 Yes (Tangential)
Byron Unit 2 19.8 0.746 0 14.8 -14.8 Yes Lower She!I Forging 49D330-1/49C298-1 19.8 1.053 5 20.8 -15.8 Yes (Axial)
Byron Unrt 2 19.8 0.746 25 14.8 10.2 Yes Lower Shell Forging 49D330-1/49C298-1 19.8 1.053 40 20.8 19.2 No (Tangential)
NOTES:
(a) Best Fit Line Per Equaten 2 of Reg. Guide 1.99 Rev. 2 Position 1.1.
Byron Dr.rt 1 Heaty, and Cooldown Umrt Curves November 1997
A9 .
.O Weld Metal:
The scatter of ARTum values (See Figure A 1) 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 O within 28'F of the best fit line. Therefore, this criteria is met for the Byron Units 1 & 2 surveillance weld material.
Base Material: ,
The scatter of ARTum values (See Figure A-1) about a best fit line drawn, as described in Regulatory Position 2.1, should be less than 17'F for base metal. As shown in Table A 3, the error for Byron Unit 1 is within 17'F of the best fit line and the error for one point for Byron Unit 2 is not within 17'F of the best fit line. Therefore, this criteria is met for Byron Unit i base metal l but not foi Byren Unit 2 base metal. As a result of the Byron Unit 2 base metal 6xceeding this criteria, the margin term that is calculated for the Byron Unit 2 base metai using Position 2.1 of the Reg. Guide should now be doubled. Thus allowing this data to be used in the evaluations of pressure-temperature limit curves and PTS.
O O
O 9
Byron Unit i Heatup and Cooldown Limit Curves November 1997
) A.10 FIGURF. A-1
) Byron l' '11 & 2 Weld Metal 80 50 ' ' ~ * ' ' * ."
an ,,,,,.........." Byron uns 1 Data a
30
s g 0
- A Dyfon WR 2 Data 20
. . . . o . One Sd Dev (28 c)
O g w % Gude .99 Equaton 2 l
' ' " . . . . . . . . . . . . . . . . . . . . . . . . . . . (7. .= .20. 6 F) ,
20 .' ,.*** ;
30 -
000E+00 $00E+18 1.00919 1.50919 2.00B19 l Fluenee, nlcm8
)
Byron Unit 1 Base Metal (SP.5933)
> 60 50 40 a Byron Wit 10sta (Armi) 30 , , , , . . . . . . . . . . . . . n. . . . . . . . . . . . . . . . . . . . . . . .
a 20 .- A Y' n N1Onta(Tang)
) -
10 "" . . . One Std Dev (17 p) 0 n ......... ...............
10 ,..***, ,.......... % Gude 1.99 Equaton 2 (G = 19 0 F) 20
) 30 0 00E+00 500E+18 1.00E+19 1.50E+19 2.00919 Fluence, n/cm8
)
)
Byron Unit i Heatup and Cooldown Limit Curves November 1997
A 11 .
.O FIGURE A 1 Byron Unit 2 Base Metal (4903301/49C2981) e 60 .
60 40 m Dyron Unit 2 Data (Amaf) 6
- u. ,,.** 3 a Byron tht 2 Data (Tang )
0 . .. . . . . One Std Dev (17 F)
, ....u........****<
10 ,,.**,...'*******,,,,,,, Reg Gude 1.99 Equaton 2 (CF e 19.8 F) $
20 30 l 0.00E+00 5.00E+18 1.00E+19 1.60E+19 2.00E+19 Fluence, nicm 8 9
Criterion 4: The Irrediation temperature of the Charpy specimens in the capsule must equal the vessel wall temperaturn at the cladding %sse metalInterface g within 4/ 25'F.
The Byron Unit 1 & 2 surveillance capsules are located in the reactor between the neutron pads and the vcssel wall ano are positioned opposite the center of the core (See Figures A 2 and A-3). The test capsules are in baskets attached to the neutron pad. The location of the 8 specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 25'F. Additionally, since the vesselinlet temperatures are the same, the irradiation tomperatures will be the same. ,
9 Dyron Unit i Heatup and Cooldown Limit Curves November 1997
)
A-12 Figure A 2: Arrangement of Surveillance Capsules in the Byron Unit 1 Reactor Vessel
)
1 l
l
) o. REACTOR VESSEL 33o' CORE BARREL Z , NEUTRON PAD CAPSULE U
)
v
- f. . . . .1. . . t.
_ ,'e i -
)
re - - - _ _
. 90' 2.5'
- l N
50' 48,7 ,
- r Y l\
X 1
)
210' iso-PLAN VIEW
)
)
)
Byron l' nit i Heatup and Cooldown Limit Curves November 1997
E A-13 .
. .o Figure A-3: Arrangement of Surveillance Capsules in the Byron Unit 2 Reactor Vessel e
O' REACTOR VESSEL CORE BARREL NEUTRON PAD (3015') Z 4 , CAPSULE U (58.5 ')
h, -z gg g, l V W *)
l 58.5' %
61' L e 270' go.
1 9
M i241 , y 7 , _J
- 1238 5') x I
g# W (121.5')
w M
3.- m q i
180' l l
eau. O A!!
ca .
>%, =,
i 7 blililllllil; i
L l ,, .
( 11 :i ss muneu no
}--- P-mm 5 52VA110N VIEW 1
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
A-14 10 -
1
, Criterion 5: The surveillance data for the correlation monitor materialin the capsule, h if present, must fall within the scatter band of the data L<se for the O material, Byron Units 1 & 2 did not incorporate correlation monitor materialin their surve'Jiance program, since this was not a requirement of E185-82. Therefore, Criterion 5 is not apphcable.
3 RESULTS & CONCLUSIONS:
3 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.
G 0
3
+
3 9
4 Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
[ ..
A 15 .
y EFFECT OF WELD METAL INTEGRATION ON BYRON P-T LIMIT CURVM:
r-Plant Previous Previous New New Results g 1/4T ART 3/4T ART il4T ART 3/4T ART Byron 1 66.37* 57.15" 70'" 60'4
- Curves at 8 EFPY FDRT/SRPLO-009(94)
Byon 2 43.5 33.2 92.6 75.8 Using weld metal Curves at integration will be more 16 EFPY restrictive to Byron 2 Pressure-temperature g
WCAP-14063 curves.
Therefore, Byron 2 curves will be regenerated and g documented in WCAP 14940 NOTES:
(a) Even after weld metalintegration, still forging-limited. Weld metalintegration has nr., effect.
(b) Calculated at 8 EFPY, (c) Calculated at 12 EFPY, g
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 documented in WCAP-14940, " Byron Unit 2 Heatup and Cooldown Curves for Normal Operation".
9 9
O Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 O
Ao16 EFFECT OF WELD METAL INTEGRATION ON BYRON PTS CALCULATIONS:
i The weld metalintegratior. CF values were calculated in Section 4 of this report. Specifically,
) the following weld metal CF values were used to determine the RTp7s values:
RG Position 1 CF RG Position 2 CF Byron Units 1 and 2 68.0'F 61.8'F
) The values listed in ' bold' below are those that were affected by the weld integration between Byron Unit 1 and Byron Unit 2. All other vessel material data was obtained from the latest PTS evaluation reportsl "". Note that for the Byron Units 1 and 2 RTprs calculations at 48 EFPY, new fluence values were interpolated to 48 EFPY The vessel surface fluence results reported in Saction 6.0 of the latest Byron Unit it'l and Byron Unit 21"1 surveillance capsule analysis reports were used.
TABLE A-4: RTp7s Values for Byron Unit 1 Material CF f*) FFN M ART,7s RT,,rs RTwot
(*F) (*F) (*F) ('F) ('F) l 32 EFPY inter. Shell Forging SP 5933 23.8 2.159 1.209 40 28.8 2 81. 8 97.6 Using sury capsule datato 19.1 2.159 1.209 40 17 23.1 80.1 I Lower Shell Forging SP 5951 26 2.159 1.209 10 31.4 31.4 72.8 t
Weld MetalWF 336 68.0 2.159 1.209 -30 56 82.2 108.2 Using surv. capsule dats'4 61.8 2.159 1.209 -30 28 74.7 72.7 48 EFPY inter. Shell Forging 5P 5933 23.8 3.238 1.309 40 31.2 31.2 102.4 Using sury. capsule data'4 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 Weld MetalWF-336 68.0 3.238 1.309 -30 56 89.0 115.0 Using surv. capsule data'4 61.8 3.238 1.309 -30 28 80.9 78.9 NOTES:
(a) 2.159 x 1019 n/cm2 (E>1.0 MeV) for 32 EFPY frorn Byron 1 PTS report (VifAP-13881). The following calculation to obtain the 48 EFPY fluence value:
2.159x1019 + (2.159x1019 - 3.807x1018)*(48 - 32 EFPY) = 3.238x1019 n/cm2 32 - 5.64 EFPY
) (b) FF (Fluence factor) = f(0.28 - 0.10* log f)
(c) Calculated using a CF based on surveillance capsule data per Regulatory Guide 1.99. Pevision 2, Position 2.
) Byron Unit i Heatup and Cooldown Limit Curves November 1997
A 17 -
TABLE A-5 RTpys VALUES FOR BYRON UNIT 2 9
Material CF ('F) f') FF5 RTuoviv M('F) ART,7s RTp1,
~
3(*F) ('F) ('F) 32EFPY Lower Shell Forging 32.2 2.192 1.213 -20 34.0 39.1 53.1 49D330 0
_-1/49C298-1 ___ . ______ ______ ______ ______ ______ ______.
Using surv. Capsule datam 19.8 2.192 1.213 -20 34'* 24.0 38.0 Inter. Shell Forging 20.0 2.192 1.213 -20 24.3 24.3 28.6 49D330-1/49C298-1 Cire. Weld MetalWF447 68.0 2.192 1.213 10 56 82.5 148.5 9 Using surv. capsule datam 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 dS33t'i4pC2991_____
Using surv. Capsule data
- 19.8 3.288 1.312 -20 34(* 26.0 40.0 l Inter. Shell Forging 20.0 3.288 1.312 -20 26.2 26.2 32.4 l 49D330-1/49C298-1 4
Cire. Weld MetalWF447 68.0 3.288 1.312 10 56 89.2 155.2
__________________ ______ ______ ______ ______ ______ _ _ _ _ _ _ .______ q Using surv. capsule data
- 61.8 3.288 10 28 81.1 113.1 1.312 l i
NOTES:
(a) 2.192 x 1019 n/cm2 (E>1.0 MeV) for 32 EFPY from Byron 2 PTS report (WCAP-14054). The following I calculation to obtain the 48 EFPY fluence value:
2.192x1019 + (2.192x101 9 3A74x1018)*(48 32 EFPY) = 3.288x1019 n/cm2 g 32 - 4.634 EFPY (b) FF (Fluence factor) = f(0.28 0.10* log f)
(c) Calculated using a CF based on surveillance capsule data per Regulatory Guide 1.99, Revision 2, Position 2.
(d) Double margin is used here due to the base metal surveillance capsule data exceeding the one sigma criteria from the credibility evaluation. See page A-9.
O t
e I
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 C
- ~
B-0
).. . .
I APPENDIX B WELD METAL INTEGRATION FOR BRAIDWOOD UNITS 1 AND 2 1
1 i
)
)
I-Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B1 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, 7
- 2. What vendor fabricated the welds and in what time frame,
- 3. What heat treatment did each surveillance program weld receive,
)
- 4. Is the initial RTc fothe welds the same or relatively close, l
S. Is the initial upper shelf energy of the welds the same or relatively close,
) 6. Is the geometry of the plants the same,
- 7. Is the type of fuel in all plants the same, 8.
) Are the fuel loading patterns 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 ternperatures do the plants operate at,
)
- 11. What are the differences in the capsule lead factors of the plants,
- 12. Can the criteria for cred2ility in 10 CFR Part 50.61 be met for each plant?
)
J J
) Byron Unit i Heatup and Cooldown Limit Curves November 1997
B2 EVALUATION:
I
- 1. What weld wire heat number, flux and flux lot numben 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).
Braidwood 2: The weld metalis 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).
The Braidwood Units 1 and 2 surveillance program weld metals were fabricated with the same heat of weld wire and the same type of flux. Therefore, this information supports the
) integration of the surveillance program test results for these welds.
- 2. What vendor fabricated the welds andin what time frame ?
Braidwood 1: B&W fabricated the welds in the late 1970's
) Braidwood 2: B&W fabricated the welds in the late 1970's The welds for Braidwood 1 and 2 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 surveillance program weld receive?
Braidwood 1: 1100 - 1150'F for 12% hours; fumace cooled.
Braidwood 2: 1150
- 50'F for 12% hours; furnace 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 the material toughness properties.
)
- 4. Is the initial RTwr of the welds the same or relatively close?
Braidwood 1: 40'F Braidwood 2: 40'F f The Braidwood Units 1 and 2 initial RTc values are identical. Therefore, this information supports the integration of the surveillance program test results for these welds.
)
) Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B-3 .
.O I
S. Is the initial upper shelf energy of the surveillance welds the same or relatively close?
l Braidwood 1: 70 ft-Ib l
Braidwood 2: 71 ft lb 8 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.
g
- 6. Is the geometry of the plants the same?
All four plants have a reactor vessel inner diameter of 173 inches, a reactor vessel beltline thickness 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 9 located at the same azimuthal angles.
- 7. Is the fuel design in allplants the same?
Braidwood 1 & 2 use 17X17 rod array fuel assemblies with the same fuel design, thus producing similar radiation effects at the surveillance capsules. O
- 8. Are the fuelloading pattems in the plants similar (i.e. Iow leakage, etc.)?
Braidwood 1 & 2 use a low leakage loading pattem.
t
- 9. What is the projected 32 effective fullpoweryear surface fluence of each plant?
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 same as Braidwood Unit 2.
O Braidwood Unit 1 0* 15' 25' 35' 45' 1.321 x10 1.984x10 2.239x10 1.86Ex10 2.162x10" 9
Braidwood Unit 2 0* 15' 25' 35' 45' 1.299x10 1.924x10" 2.199x10 1.861x10 2.174x10
4 4
Byron Unit 1 Heatup anc Cooldown Limit Curves November 1997
B4 I.
- 10. What vesselinlet temperatures do the plants operate? (Per Reference 25)
CYCLE # Braidwood Unit 1 Tcold (*F) Braidwood Unit 2 Tcold ('F) 1 557 (Cap. U)"* 557 (Cap. U)"*
2 551 551 3 551 551 4 551 (Cap. X)"* 551 (Cap. X)"*
5* 551 551
) 5" 554 --
6 554 551 Between A1R04 & A1M05 (Approximately % cycle duration)- Unit 1 Only.
)
" Between A1M05 & A1R05 (Approximately % cycle duration)- Unit 1 Only.
"* See Appendix C, page C-5.
- 11. What are the differences in the capsule lead factors of the plants?
Based on the information provide in Table B-1, the lead factors of the surveillance capsules
)
in Braidwood Unit 1 are essentially the same as Braidwood Unit 2.
TABLE B-1 Surveillance Capsule Lead Factors for Braidwood Units 1 & 2 i
Braidwood Unit 1 Braidwood Unit 2 Capsule Location Lead Factor Capsule Location Lead Factor U 58.5* 4.03 U 58.5* 4.00
)
X 238.5a 4.03 X 238.So 4.02 W 121.So 4.03 W 121.5 4.02
. Z 301.5 4.03 Z 301.So 4.02
)
V 61.0o 3.73 V 61.0 3.70 Y 241.0o 3.73 Y 241.0o 3.70
)
Based on the projected vessel surf ce fluence and lead factor values for Braidwood 1 & 2, the Braidwood 1 & 2 surveillance capsules will have approximately the same flux rates and
) Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B5 .
.O irradiation temperatures. This supports the use of the surveillance weld data in ooth programs to evaluate the reactor vesselintegrity of the Braidwood units.
5
- 12. Can the criteria for credibility in 10 CFR Part 50.61 be met for each plant?
Credibility will be evaluated for all the surveillance capsule data (base metal & weld metal) for Braidwood Unitt,1 and 2.
A Criterion 1: The materials in the surveillance capsules must be those which are controlling materials with regard to radiation embrittlement.
The following is a list of the beltline inaterials contained in the Braidwood Units 1 and 2 surveillance programs: S 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.) I 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 8061. (This is the same heat number used g in the limiting beltline weld.)
Based on the information provided in the material selection documents, WCAP-9807 (Braidwood 1, See Ref. 26) and WCAP-11188 (Braidwood 2, See Ref. 27), these materials are #
judged to be the most controlling with regard to radiation embrittlement for each unit. Therefore, Criteria #1 is met for both Braidwood units.
Criterion 2: Scatterin the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions must be small enough to permit the determination of the 30 ft Ib temperature unambiguously.
Plots of Charpy energy versus temperature for the unirradiated condition are presented in WCAP-9807, " Commonwealth Edison Company Braidwood Station Unit No.1 Reactor Vessel 4 Radiation Surveillance Program," dated February 1981 and WCAP-11188, " Commonwealth Edison Company Braidwood Station Unit No. 2 Reactor Vessel Radiation Surveillance 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 units.
{
Byron Unit i Heatup and Cooldown Limit Curves November 1997 C
f B6 Based on engineering judgement, the scatter in the data presented in these reports is small enough to determine the 30 ft-Ib temperature and the upper shelf energy of the Braidwood Units 1 & 2 surveillance weld metals unambiguously. Therefore, the Braidwood Units 1 & 2 h surveillance materials meet this criteria.
Criterion 3: Where there are two or more sets of surveillance data from one reactor, the scatter of ARTer values must be less than 28 F for welds and 17 F for y base metal Even if the range in the capsule fluences is large (two or more orders of magnitude), the scatter may not exceed twice those values.
The least squaret 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. It should be noted here that the ratio procedure is not applied in this instance since only surveillance capsule data is being evaluated.
)
)
)
)
7
) Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B-7 .
TABLE B-2*
Braidwood Units 1 & 2 Surveillance Capsule Data Chemistry Factor for Best Fit Line 9
Material Capst!e f* FFW Measured FF x FF8 ARTwot ARTwo, Braidwood Unit 1 0 3.814x10ia 0.733 5 3.666 0.538 Lower Shell Forging 8 490867-1/49C813-1 X 1.144x10" 1.038 30 31.127 1.077 (Axial)
Braidwood Unit 1 U 3.814x10" O.733 0 0.000 0.538 Lower Shell Forging X 1.144x10" 1.038 25 25.939 1.077 I 49D867-1/49C813-1 Sum: 60.733 3.228 (Tangential)
Chemistry Factor = 60.733 + 3.228 = 18.8'F l Braidwood Unit 2 U 3.933x10 O.741 0 0.00 0.550 $
Lower Sheli Forging 50D102-1/50C97-1 X 1.126x10 1.033 3 3.099 1.067 (Axial)
Braidwood Unit 2 U 3 933x10 O 741 5 3.707 0.550 g
Lower Shell Forgin9 X 1,126x10 35 1 033 36.160 1.067 50D102-1/50C971 Sum: 42.966 3.234 (Tangential)
Chemistry Factor = 42.966 + 3.234 = 13.3'F Braidwood Unit 1 U 3.814x10 O.733 10 7.333 0.538 Weld Metar') X 1.144x10 1.038 25 25.95 1.077 Braidwood Unit 2 U 3.933x10 O.741 0 0.00 0.550 Weld Metaf*) X 1.126x10 1.033 20 20.66 1.067 &
Sum: 53.943 3.232 Chemistry Factor = 53.943 + 3.232 = 16.7'F NOTES:
(a) f = Fluence (1019 n/cm2, E > 1.0 MeV)
(b) FF = Fluence Factor . f(0.28 - 0.1
- log f) 4 (c) Values of f, FF, and ARTNOT values were taken from Table 2 of WCAP-14243 (Braidwood Unit 1 P-T Limits) and WCAP-14230 (Braidwood Unit 2 P-T Limits).
(d) CF = I(FF*RTNDT) + E(FF2)
(e) For both welds: WF-562, heat # 442011.
l Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
v B-8 I l
I TABLE B-3 Best Fit Evaluation for Braidwood 1 & 2 Surveillance Materials Base Material CF FF ART,.y Best Fit Scatter of < 17'F (Base Metals)
(30 ft-Ib) (*F) ART,.y (*F) ART,er (*F) < 28'F (Weld Mete!)
16.7 0.733 10 12.2 -2.2 Yes Braidwood 1 Weld Metal 1S.7 1.038 25 17.3 7.7 Yes 16.7 0.741 0 12.4 -12.4 Yes Braidwood 2 Weld Metal 16.7 1.033 20 12.3 7.7 Yes Braidwood Unit 1 18.8 0.733 5 13.8 -8.8 Yes Lower Shell Forging 49D867-1/49C813-1 18.8 1.038 30 19.5 10.5 Yes (Axial)
Braidwood Unit 1 18.8 0.733 0 13.8 -13.8 Yes Lower Shell Forging 49D887-1/49C813-1 18.8 1.038 25 19.5 5.5 Yes (Tangential)
Braidwood Unit 2 13.3 0.741 0 9.9 -9.9 Yes Lower Shell Forging 50D102-1i50C97-1 13.3 1.033 3 13.7 -10.7 Yes (Axial)
Braidwood Unit 2 13.3 0.741 5 9.9 -4.9 Yes Lower She3 Forging 50D102-1/50C97-1 13.3 1.033 35 13.7 21.3 No (Tangential)
NOTES:
(a) Best Fit Line Per Equation 2 of Reg. Guide 1.99 Rev. 2 Position 1.1.
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 i
l
B9 '
.C Weld Metal:
The scatter of ARTum values (Figure B-1) 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 W of the best fit line. Therefore, this criteria is met for the Braidwood Units 1 & 2 surveillance weld material.
Base Material:
The scatter of ART,m values (Figure B-1) about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 17'F for base metal. As shown in Table B-3, the error for Braidwood Unit 1 is within 17'F of the best-fit line and the error for one point for Braidwood Unit 2 is not within 17'F of the best-fit line. Therefore, this criteria is met for Braidwood Unit 1 base W metal but not for Braidwood Unit 2 base metal. As a result of the Braidwood Unit 2 base metal exceeding this criteria, the margin term that is calculated for the Braidwood Unit 2 base metal using Position 2.1 of the Reg. Guide should now be doubled. Thus aliowing this data to be used in the evaluations of pressure-temperature limit curves and PTS.
5 e
e 4
e f
Byron tinit 1 Heatup and Cooldown Limit Curves November 1997 Ol
B 10
),
Figure B 1 7 Braidwood Units 1 & 2 Weld Metal 80 .
50 40 g Bradw ood Unit 1 Data
) 30
- u. 8 A Bradw ood und 2 Data 4 20 a -_
g ', --g A
. . . . . . . one Se D. <2. n Reg Guide 1.99 Equaton 2 10
) ........-..........."--. I ger ,3,7 p) 20 30 0.00E+00 5.00 &18 1.00E+19 1.50E+19 2.00E+19 <
Fluenee, nicm 8
)
)
l Braidwood Unit 1 Base Metal (49D867-1149C813-1)
) 80 50 43 m Bradw ood und 1 Data
,,,,,,,,.. ......" " '.- -- (Axial)
- u. "
.. **** Bradw ood thit 10sta A
4 20 - (Tang) 10 . . . . . . . One Std Dev (17 F) s, .
^
10
- .... ~ *** " " ............... Reg Guide 1.99 Equaten 2 (CF = 18.8 F) 20 .
30 0.00E+00 5.00E+18 - 1.00E+19 1.50E+19 2.00E+19 Fluence, nicm 2 2
) Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B-11 ,
,C Figure B-1 (Continued) i Braidwood Unit 2 Base Metal (50D102-1/50C97-1) g 60 50 40 m Bradw ood Unit 2 Data
(
30 * * * * * * * " O
- u. Bradw ood Unit 2 Data 4 20 ,*,,,....-****"...... 3 (Tang) 10 . . . . . . . One Std Dev (17 F) z o n
- ,,,,,,,.........-...".*"**"*" Reg Guide 1.99 Equabon 2 10 ,,... (cF = 13.3 F) g 20 .
30 i 0.00&OO 5.00618 1.00 & 19 1.50619 2.00919 Flue nee, nicm 8 9
l l
l Criterion 4: The Irradiation temperature of the Charpy specimens in the capsule must equal the vessel wa!! temperature at the cladding / base metal ,
Interface within +/- 25 F.
The Brsidwood Unit 1 & 2 surveillance capsules are located in the reactor between the neutron pads and the vessel wall and are positioned opposite the center of the core (See Figures B-2 and B-3). The test capsules are in baskets attached to the neutron pad. The location of the 81 specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 25'F. Additionally, since the vessel inlet temperatures are the same, the irradiation temperatures will be the same.
4 l
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 O
} B-12 Figure B-2: Arrangement of Surveillance Capsules in the Braidwood Unit 1 Reactor Vessel
)
C' f" REACTOR VESSEL
)
g CORE sARRn
--umrRou pao (301 5') Z 7
, CAPSULE U (58.5')
J v (ei'i
)
'- / 58.5' %
gge 270* Ir -
go.
L (241 ') y I I (238 5') x W (121.5')
l REActon VESSEL 1so' I
= view !
! p vEssa sf WAu.
) ) COREhi; /
C# WLE ASSEMSLY lilllllllli i
{s CORE
'! - s' W LANE I l l N NEUTRON PAD t is i E
- CORE BARREL ELEVATION VIEW
)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B 13
- Figure B-3: Arrangement of Surveillance Capsules in the Braidwood Unit 2 Reactc c Vessel O'
REACTOR VESSEL
,7 CORE BARREL NEUTRON PAD (301. 5 ') Z '
CAPSULE U (58.5 ')
58.5, (6 )
,G8.5 ' % _
s.
270* .2 .
90' p
(241') y '-
(238.5') X I Y -
W (121.5 ')
/
REACTOR WME O 180* I PLAN VIEW -
e f VESSEL
[ WALL g
! s CAPSULE ELEVATION VIEW CORE : NF ASSEMBLY llIIIllllll! $
l ;
d r.- : :
E s
N
, _.- CORE MIDPLANE NN NEUTRON PAD 1
1 :M
" CORE BARREL l
Byron Ur.it 1 Heatap and Cooldown Limit Curves a November 1997
B-14
).
Criterion 5: The surveillance data for the correlation monitor materialIn the capsule, if present, must fall within the scatter band of the data base
) for the m:sterial.
Braldwood Units 1 & 2 did not incorporate correlation monitor material in their surveillance program, since this was not a requirement of E185-82. 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 Braidwood Unit 1 wela metal surveillance data with Braidwood Unit 2 weld metal surveillance data.
I
) '
)
)
)
)
l
)
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997 i
3
_ B 15 ,
TABLE B-4 Calculation of Average Cu and Ni Weight Percent Values for the Braidwood Weld Mat 5 rial (Using Braidwood 1 & 2 Chemistry Test Results) 9 Best-Estimate Reference Cu Ni B&W Weld Qualification BAW-2261 0 028 0.63 B&WWeld Ouakfication 0.03 0.65 M W We'a Ot.alification 0.04 0.67 Braidwood 1 Surv. Data See below 0.032 0.671 -> 0.03 0.67 g
Surv. CF = 41 P.T.idwriod 2 Su v. Data See Below 0.033 0.708 > 0.03 0.71 Surv. CF = 41 Best-Et.timate 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 Surveillance Chemistry Results:
Braldwood Unit 1 Braldwood Unit 2 Reference Cu N1 Reference Cu Ni WL AP-9807 0.04* 0.67* WCAP-11188 0.040 0.64 WCAP-12685 0.035 0.666 WCAP 14228 0.033 0.724 0.033 0.666 0.034 0 711 I 0.034 0.723 0.033 0.714 0.035 0.709 0.038 0.780 0 034 0.728 0.035 0.737 0.035 0.699 0.033 0.728 0.035 0.751 0.032 0.752 0.031 0.683 0.032 0.743 g 0.032 0.673 0.031 0.730 0.029 0.666 0.032 0.711 0.029 0.686 0.032 0.728 0.034 0.610 0.031 0.703 0.033 w.651 0.032 0.687 0.033 0.698 0.033 0.703 0.031 0.656 0.033 0.695 O WCAP 14741 0.031 0.635 WCAP-12845 0.032 0.704 0.029 0.647 0.034 0.754 0.028 0.638 0.032 0.698 0.031 0.655 0.026 0.623 0.031 0.650 0.028 0.635 0.032 0.661 0.031 0.679 0.033 0.667 g
0.029 0.644 0.028 0.648 0.032 0.699 0.0; ' O.644 0.034 0.765 0.0'>4 0.668 0.031 0.673 0.033 0.656 0.034 0.724 0.036 0.658 0.035 0.747 0.036 0.671 0.036 0.667 0.033 0.711 $~
0.031 0.688 Averap 0.032 0.671 0 03'i 0750 0.031 0.685 Average 0.033 0.708 Not used in Average calculation; reported for completeness. The same value appears in the material test reports and the surveillance program report.
" The best estimate chemistry values was obtained using the " average of averages" approach.
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
) B-16 EFFECT OF WELD METAL INTEGRt. TION ON BRAIDWOOD P T LIMIT CURVES:
) Plant Previous Previous New New Result 1/4T ART 3/4T ART 1/4T ART 3/4T ART Braidwood i 76.6 65.4 69.7 60.6 Current curves / PTS Curves at evaluation are 16 EFPY conservative.
WCAP44243 New Applicability Date:
27.9 EFPY '
Braidwood 2 62.6 55.7 69.5 60.4 Current curves / PTS Curves at evaluation are NOT 16 EFPY
) conservative. Using weld metalintegration wil! bo WCAP 14230 more rsstrictive.
New Applicability Date:
7.4 EFPY
) Braidwood 2 curves will be
' regenerated and documented in WCAP-14970 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 for Braidwood Unit 1.
) 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:
- FF)
Using the " Previous" ART values and initial rte, this equation was used to back-calculate the fluence factor (FF) and the vessel surface fluence value. This fluence value was then used to I determine a new applicability date (in terms of EFPY) for the current pressure-temperature limit curves.
For Braidwood Units 1 and 2, the margin 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 Guide 1.99, Revision 2. A more detailed explanation can be found in Section 4 of this report.
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B 17 .
.C The Margin term is calculated as, M = 2do;' + cn2 . The standard deviation for the initial RTuor margin term (c) is 0 F when the initial RTuor is a measured value (as is the case for the Braidwood units). Additionally, the term og need not exceed 0.5 times the mean value of g ARTuor.
Therefore, when the ARTuor value is mdtiplied by 0.5 and plugged into the above equation, the effect is 2 * (ARTuny /2), which is the ARTuor (or CF
- FF).
- FF) 76.6'F = 40*F + (16.7
- 1/4T FF)*F + (16.7
- 1/4T FF)*F = *> 1/4T FF = 1.0958 g 1.0958 = 1/4T fars.u w v4ro ==> 1/4T f = 1.4124 x 10 n/cm2 1.4124 x 10 = f
- eH 2' * " * "5) = = > f = 2.352 x 10 n/cm2 W
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 h
2.352 x 10 = 2.239 x 10 + (X - 32 EFPY) * (2.239 x 10- 1.120 x id')
Weld Metal calculations based on a 3/4T ART = 65.4 F:
(The following data is from Braidwood Unit 1 heatup/cooldown curve report, WCAP-14243) 8 ART = 1 + M + (CF
- FF) 65.4 F = 40 F + (16.7
- 3/4T FF)*F + (16.7
0.76047 = 3/4T f( 2s.o w wro ==> 3/4T f = 0.4221 x 10 n/cm2 0.4221 x 10 = f
- e*2"'"5) ==> f = 1.9495 x 10 n/cm2 This fluence value will occur between 16 and 32 EFPY, per Table 6-15 of WCAP-14241. The following calculation will determine the applicability date in terms of EFPY.
Byron Unit i Heatup and Cooldown Limit Curves November 1997
. B-18 Fluence at X EFPY = Fluence at 16 EFPY + (X - 16 EFPY)
- Fluence /EFPY O 1.9493 x 10 = 1.120 x 10 + (X - 16 EFPY) * (2.239 x 10- 1.120 x 10')
) Therefore, after the weld metal integration for Braidwood Units 1 and 2 is implemented, the Braidwood Unit i heatup/cooldown curves presented in WCAP-14243 will be applicable to 27.9 EFPY, D
BRAIDWOOD UNIT 2 l Weld Metal calculations based on a 1/4T ART = 62.6 F:
O (The following data is from Braidwood Unit 2 heatup/cooldown curve report, WCAP-14230.)
- FF) g 62.6*F = 40*F + (16.7
- 1/4T FF)'F + (16.7
. 3.075 x 10 = f
- e*8"''25) ==> f = 5.120 x 10 n/cm2 O
This fluence value wi:1 cccur 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.
g 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 108) 16 - 4.215 EFPY X = 7.4 EFPY 4
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B-19 .
C I
WeH Metal calculations based on a 3/4T ART = 55.7'F:
- FF) U 55.7'F = 40*F + (16.7
- 3/4T FF)"F + (16.7
0.1292 x 10 = f
- ew24 e s no ==> f = 5.966 x 10 n/cm' This fluence value 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. 9 Fluence at X EFPY = Fluence 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- 2.896 x id') ,
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 applicable to 7.4 Ei "'Y. g 6
9 9
O
(
Byron Unit 1 Heatup and Cooldown Limit Curves Novembe* 1997 dl
B 20 0,
EFFECT OF WELD METAL INTEGRATION ON BRAIDWOOD PTS CALCULATIONS:
The weld metal integration eF values were calculated in Section 4 of this report. Specifically, Q the following weld metal CF values were used to determine the RTers values: l RG Positiom 1 CF RG Position 2 CF Braidwood Units 1 and 2 41.0*F 16.7'F The values listed in ' bold' below are those that were affected by the weld integration between l Graidwood Unit 1 and Braidwood Unit 2. All other vessel material data was obtained from the latest PTS evaluation reportsne .2oi, P
l TABLE B-4 l
RTers Values for Braidwood Unit 1 Material CF f FF* M ARTry RTuorcui RTris 3 ('F) (*F) ('F)
{.p) .p) 32 EFPY I
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 31.63 31.68 43.4 using S/C data *) 18.8 2.239 1.218 -20 1? 22.90 19.8 Weld MetalWF 562 41.0 2.239 1.218 40 49.95 49.95 139.9 using S/C data *> 16.7 2.239 1.218 40 20.34 20.34 80.7 3 48 EFPY Inter. Shell Forging 31.0 3.358 1.317 -30 34 40.83 44.8 Lower Shell Forging 26.0 3.358 1.317 -20 34 34.25 48.3 g using S/C data *> 18.8 3.358 1.317 20 17 24.76 21.8 Weld Metal WF-562 41.0 3.358 1.317 40 54.00 54.00 148.0 using S/C data *) 16.7 3.358 1.317 40 21.99 21.99 84.0 NOTES:
(a) FF (Fluence factor) = f(0.28 - 0.10* log f)
(b) Calculated using a CF based on surveillance capsule data per Regulatory Guide 1.99, Revision 2. Position 2.
?
7~
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
B-21 ,
.h TABLE B-5 RTns Values for Braidwood Unit 2 Material CF ('F) f FFio RTuoyv3 M('F) ART,13 RTp1, C
('F) ('F) op) 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 34t" 16.15 20.2 Weld Metal WF-562 41.0 2.199 1.214 40 49.77 49.77 139.5 using S/C datam 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 l using S/C data *) 13.3 3.298 1.313 30 34t4 17.46 21.5 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 NOTES:
(a) FF (Fluence factor) = f(0.28 0.10* log f)
(b) Calculated using a CF based n surveillance capsule data per Regulatory Guide 1.99, Revision 2. Position 2.
y (c) Double margin is used here due to the base metal surveillance capsule data exceeding the one sigma criteria from the credibility evaluation. See page B-9.
1 i
e D
l I
l l
Byron Unit 1 Heatup and Cooldown Limit Curves November 1997
C-0
)
APPENDIX C BYRON /GRAIDWOOD FLUENCE METHODOLOGY JUSTlFICATION AND TIME-DEPENDENT CAPSULE FLUENCE VALUES l
l 2
)
)
)
?
)
syron unit 1 Heatup anc vooicown uma curves Novemoer 1m
1, C1 1 Fluence Methodology Justificatinn
)
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 Curve:,"is consistent with the requirements of Draft Regulatory Guide DG 1053,
" Calculational and Dosimetry Methods for Determining Pressure Vessel iJeutron Fluence" and
) makes use of neutron transport cross scetions derived from the ENDF/B VI dait base. The exposure evaluations documented in WCAPs 13880,14064,14241, and 14228 for the Byron Units 1 & 2 gnd Braidwood Units 1 & 2 pressure vessels were completed prior to the release of the ENFfB VI based Light Water Reactor neutron transport cross section library, Consequently the neutron transport calculations r.erformed 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 applied to the Byron Units 1 & 2 and Braidwood Units 1 & 2 fluence evaluations was identical to the approved methodology described in WCAP 14040 NP A.
I Cornmonwealth Edison plans to re-evaluate capsule and vessel fluence estimates utilizing ENDF/B Vi neutron cross section libreries in accordance with WCAP 14040 NP A at the next scheduled capsule Mdr1wal for each set of units (capsule W for Byron Unit 1 at BIR08 in November 1997 anc ~psule W for Braidwood Unit 1 at AIR 07 in September 1998), and for all subsequent caosule withdrawals, and proposes to integrate data pursuant to 10 CFR 50 Appendix H. This will replace estimates previously performed using a combination of ENDF/B-IV transport cross sections and ENDF/B-V dosimetry cross sections. Since this re-evaluation willimpact ine manner in which materials data are utilized, and, therefore constitutes a change in PTLR methodologies, all revised values of rat resulting from the new fluence values along
) with an ev.Justion of their impact on pressure temperature limits will be submitted to NRC for review and approval, in addhion to the methoc' ology upgrade discussed in the preceding paragraph, the fluence upcates for Byron tsnits 1 & 2 and Braidwood Units 1 & 2 will also include an evaluauan of low j
leakage fut1 management instituted at all four units A qualitative excmination of the loading pattems ucad at Byron Units 1 & 2 and Braidwood Units 1 & 2 indicates that accounting for the flux reduction 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 net effect of methods upgrades and low leakage fuel management on 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 14220 to the approved methodology described in WCAP-14040-NP-A, including the impact of low leakage fuel management, coupled with the low sensitivity to irradiation damage exhibited by the materials uyron unn 1 rieawp ana vooicown uma uurves riovemoer 1m 7 w ~
C2 .
,C comprising the Byron Units 1 & 2 and Braidwood Units 1 & 2 reactor pressure vessels, the use of the previously documented fluence values is justified until the update to the ENDF/B VI based methodology is completed for each unit. 9 2 Time Dependent Surveillance Capsule Fluences O
Based on the documentation provided in WCAPs 13880,14064,14241, anc 14228, it 16 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 surveillanets capsule locations for times beyond those withdrawal dates are needed g in order to estabil.h approprkite withdrawal schedules for the remaining capsules comprising the Reactor Vessel Surveillance Program for each of the units. These Best Estimate projections are piovided in Tables C-1 through C 4 for Byron Unnt 1 & 2 and Braidwood Units 1
& 2, respectively. These projections are based on the assumption that the best estimate j neutron flux averaged over the total irradiation time for each unit would remain applicable for the e remainder of plant lifetime.
3 Comparison of Irradiation Environments D
I Byron Units 1 and 2 as well as Braidwood Units 1 and 2 are Westinghoure reactors employing a reactor internals design using partially circumferential neutron pads. The surveillance capsules holding the materials test specimens are mLnted on the outer radius of the neutron pads in the downcomer region between the pressure vessel wall and the thermal shield. Thus, the surveillance specimens are mounted behind the full thickness of the neutron pad. The 8 location of the maximum fluence on each of the respective pressure vessels also occurs ai in azimuthal angle behind the neutron pad. The geometry of the neutron pads, surveillance capsules and associated support structure, and the pressure vesselitself are all modeled explicitly in the neutron transport calculations performed for the Byron and Braidwood reactors.
The design of the Byron and Braidwood reactor internals includes former plates at several axial intervals spanning the radial distance between the external boundary of the baffle plates and the inner radius of the core barrel. Due to the shape of the perimeter of the reactor core, the radial extent of the former plates varies significantly with azimuthal angle. The presence of g these former plates can have a localized effect on any dosimetry or meterials test specimens that may be located directly in line with these steel plates. Studies have been performed to estimate the effect of the former plates on the irradiation conditions within the surveillance capsules. The results of these studies indicate that the formers have the largest effect on high threshold reactions such as Cu (n,a), Fe (n,p), and Ni (n,p) and a minimal effect on the lower ,
threshold reactions and exposure parameters such as U-238 (n,f), Np-237 (n,f), Fluence (E >
1.0 MeV), and dpa. The maximum effects noted were 11% for Cu,10% for Fe,9% for Ni,7%
uyron una 1 neatup ana vooicown umit uurves r9ovemoer ww O
C-3 1.0 MeV), and dpa. The maximum effects noted were 11% for Cu,10% for Fe,9% for Ni,7%
for U 238,1% for Np 237,3% for Fluence (E > 1.0 MeV), and 1% for dpa. Each of these
) percentages represent a reduction in the calculated value at a location directly in line with the formerplates. In the case of the Byron and Braidwood units, neutron dosimeters are dispersed axially within the capsules such that the effects introduced by the presence of the former plates are minimized.
? Since these four reactors were designed as identical units, the plant specific geometries in the vicinity of the surveillance capsules tend to result in radiation environments at the capsule l positions that are almost identical both quantitatively and qualitatively. In all units, the capsules are designed to minimize the impact of gamma ray heating and, thus, maintain the irradiation temperature of the test specimens close to the temperature of the coolant in the downcomer region. Likewise, the temperature of the pressure vessel wall at the clad / base metalinterface is also maintained very close to the downcomer coolant temperature, thus, providing compatibility between the test specimen irradiation temperatures and that of the pres =ure vessel wall,
) To date eight surveillance capsules have been withdrawn from the same symmetric 31.5' azimuthallocation at the four Byron and Braidwood reactors (two from each unit). Comparisons of the neutron dosimetry evaluations performed for these 8 surveillance capsules provide an indication of the similarity in the irradiation environments for each of these capsules, h
- a) Damage Rate I
The following tabulation provides the neutron flux (E > 1.0 MeV) and iron atom displacement rate (dpa/sec) experienced by each of the surveillance capsuies withdrawn from Buron Units 1 ar.d 2 and Braidwood Units 1 and 2. The damage rates represent an average over the total
] irradiation period experienced by the respective capsules.
Flux (E > 1.0 MeV) Displacement Rate Irradiation Time Capsule [n/cm'sec) [dpa/sec) [efgy)
Byron 1 - U 9.86e+10 1.89e-10 1.15 3 Byron 2 - U 1.10e+11 2.06e 10 1.15 Braidwood 1 U 1.10e+11 2.10e-10 1.10 Braidwood 2 - U 1.08e+11 2.07e 10 1.15 Byron 1 - X 8.11e+10 1.55e-10 5.64 Byron 2 W 8.27e+10
) Braidwood 1 - X 8.5?e+10 1.53e-10 1.57e-10 4.64 4.23 Braidwood 2 X 8.46e+10 1.56e-10 4.22
) An examination of this tabulation shows that in terms of neutron flux (E > 1.0 MeV) the range of damage rate extends from 8.11e+10 to 1.10e+11 and in terms of iron atom displacement rates the correspondin0 range extends from 1.53e 10 to 2.10e 10. The total range of damage rates
) uyton unit 1 Heatup ana vooloown umit vurves Novem>er wW
,C for these capsules falls within approximately a factor of 1.4. Furthermore, there is no systematic difference among any of the units.
9 b) Spectral Balance
/.n indication of the differences in the energy distribution of neutrons at the surveillance capsule locations can be obtained via a comparison of the ratio of [dpa/sec)/[ flux (E > 1.0 MeV)) at the respective capsule locations. A comparison of this type for the Byron and Brak od O surveillance capsules is provided as follows:
Irradiation Time Capsule [dpa/ Flux) [efpy)
Byron 1 - U 1.92e 21 1.15 Byron 2 - U 1.87e 21 1.15 Braidwood 1 - U 1.91e 21 1.10 Braidwood 2 - U 1,92e 21 1.15
- Byron 1 - X 1.91e-21 5.64 l
a Byron 2 - W 1.85e 21 4.64 Braidwood 1 X 1.83e-21 4.23 Braidwood 2 - X 1.84e-21 4.22 An examination of this data table shows that the spectralindices as expressed by the ratio of D iron displacement rate to neutron flux (E > 1.0 MeV) varies by less than approximately 5% over the entire range of capsules included in the data set. Thus, from a spectrum balance viewpoint the Byron and Braidwood irradiation conditions are essentially identical.
c) gamma heating A Gamma heating effects the irradiation environment of the Byron and Braidwood reactor vessels in a similar fashion. At the surveillance capsule and reactor vessellocations the gamma ray heating is due primarily to secondary gamma rays induced by inelastic scattering and neutron capture in local materials. Since the secondary gamma ray production is a function of the neutron energy spectrum which, as described above, is essentially identical for these reactors, it A follows that the gamma ray heating for these reactors is also very similar.
Furthermore, in the design of the Westinghouse surveillance capsules, the impact of internal heat generation is small and the specimen temperatures are maintained very close to the g downcomer water temperature.
9 t$yron Unit 1 tieatup ana Uoolaown umit vurves riovemoeTTW7 .I
),
C5 d) Irradiation Temperature
(
) The vesselinlet temperatures for the Byron and Braidwood units for each of their operating fuel I cycles are listed as follows:
Fuel Cycle Byron i Byron 2 Braidwood 1 Braidwood 2
) 1 557 557 557 557 2 557 551 551 551 3 551 551 551 551 '
4 551 551 551 551 5 551 551 551 554 551
) 6 551 551 554 551 7 561 551 8 551 Note: Byron Units 1 & 2, and Braidwood Units 1 & 2 capsules U were withdrawn at the end of i
cycle 1. Byron 1 capsule X was withdrawn at the end of cycle 5 and Byron 2 capsule W was
! withdrawn at the end of cycle 4. Braidwood 1 & 2 capsules X were both withdrawn at the end of cycle 4.
)
! For Braidwood Units 1 and 2, the vessel inlet temperatures were identical from one unit to the other during the cycles that the capsules in question were being irradiated. Thus, the irradition conditions were identical.
) For Byron Units 1 and 2, on the other brnd, there exists a 6'F temperature difference in cycle
- 2. Since Byron 2 capsule W was pulled at the end of the fourth cycle, the time weighted temperature difference over the four cycles that the capsules (i.e. Capsules X and W) were irradiated is 1.5'F, or conservatively 2'F. For the purpose of evaluating surveillance weld data credibility in Tab! A-4 and for the purpose of evaluating weld chemistry factor normalized to the
) Byron Unit 1 and Byron Unit 2 reactor vessels in Table 4 (page 11), the effects of this adjustment to the measured ARTuor by 2'F (conservatively applied to raising the lower temperature Byron 2 capsule W measured ARTuor)is very small. The effect of the 2*F adjustment , and even the effect of applying the entire sii gle cycle O'F temperature difference to
) the Byron 2 capsule W measured ARTuot, is completely compensated for with the application of the conservative ratio of 3.0, relative to the actual ratio of 2.5. Temperature difference and chemistry factor ratios will be re-evaluated at all future scheduled capsule evaluations.
T
) tsyron una 1 tieatup ana voolaown uma curves reovemoer im
C6 .
8 Table C 1 9
BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS AT SURVEILLANCE CAPSULE LOCATIONS BYRON UNIT 1 Irradiation Fluence [n/cm') Lead Factor Time
[EFPY) 31.5 Caps 29.0 C1 31.5 Caps 29.0 Caps 5.64 1.443e+19 1.365M 1 3.79 3.58 8 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 18.00 4.604e+19 4.353e+19 3.79 3.58 g 20.00 5.116e+19 4.837e+19 3.79 3.58 22.00 5.628e+19 5.321e+19 3.79 3.58 24.00 6.139e+19 5.804e+19 3.79 3.58 4
26.00 6.651e+19 6.288e+19 3.79 3.58 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 %
6 tsyron ung 1 Heatup ana voolaown uma curves Novemoer ww
C7
).
Table C 2
)
BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS AT SURVEILLANCE CAPSULE LOCATIONS - BYRON UNIT 2 I
) Irradiation Fluence [n/cm'] Lead Factor l Time
[EFPY] 31.5 Caps 29.0 Caps 31.5 Caps 29.0 Caps
) 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 j 10.00 2.664e+19 2.491e+19 3.89 3.64 12.00 3.197e+19 2.989e+19 ?.89 3.64 14.00 3.730e+19 3.487e+19 3.89 3.64 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 3
24.00 6.394e+19 5.977e+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 3.64
)
) -
cyron unit 1 tiestup ana voooown umit vurves novemoer avw
C ,
.v Table C 3 T
BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS AT SURVEILLANCE CAPSULE LOCATIONS BRAIDWOOD UNIT 1 Irradiation Fluence [n/cm'] Lead Factor Time
[EFPY] 31.5 Caps 29.0 Caps 31.5 Caps 29.0 Caps 4.23 1.193e+19 1.105e+19 4.02 3.73 0 6.00 1.690e+19 1.565e+19 4.02 3.73 8.00 2.254e+19 2.087e+19 4.02 3.73 10.00 2.817e+19 2.609e+19 4.02 3.73 &
12.00 3.380e+19 3.130e+19 4.02 3.73 14.00 3.944e+19 3.652e+19 4.02 3.73 16.00 4.507e+19 4.174e+19 4.02 3.73 g 18.00 5.070e+19 4.696e+19 4.02 3.73 20.00 5.634e+19 5.217e+19 4.02 3.73 22.00 6.197e+19 5.739e+19 4.02 3.73 6
24.00 6.761e+19 6.261e+19 4.02 3.73 26.00 7.324e+19 6.783e+19 4.02 3.73 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 8.34Ce+19 4.02 3.73 6
byron una 1 Heatup ana cooloown umit vurveS C
NovemDef TWW(
C9
).
Table C-4
)
BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS AT SURVEILLANCE CAPSULE LOCATIONS BRAIDWOOD UNIT 2 h Irradiation Fluence [n/cm'] Lead Factor l Dme
[EFPY1 31.5 Caps 29.0 Caps 31.5 Caps 29.0 Caps
) 4.21 1.163e+19 1.072e+19 4.02 3.70 6.00 1.656e+19 1.526e+19 4.02 3.70 8.00 2.208e+19 2.034e+19 4.02 3.70
) 10.00 2.760e+19 2.543e+19 4.02 3.70 12.00 3.312e+19 3.051e+19 4.02 3.70 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+1b 4.02 3.70 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 8.281e+19 7.628e+19 4.02 3.70 32.00 8.833e+19 8.136e+19 4.02 3.70
)'
y tsyron unn 1 r1eamp ana voonsown uma curves novemoer lvw
_ . _ _ _ _ _ _ _ _ - -