ML20141A468

Rev 1 to Byron Unit 1 Heatup & Cooldown Limit Curves for Normal Operation & Surveillance Weld Metal Integration for Byron & Braidwood

ML20141A468
Person / Time
Site:	Byron, Braidwood
Issue date:	04/30/1997
From:	Grendys P WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20141A461	List: ML20141A459 ML20141A468
References
WCAP-14824, WCAP-14824-R01, WCAP-14824-R1, NUDOCS 9705140323
Download: ML20141A468 (64)

Text

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BYRON UNIT 1  !

l HEATUP AND l COOLDOWN LIMIT CURVES FOR .

NORMAL OPERATION l AND SURVEILLANCE .

i WELD METAL INTEGRATION FOR BYRON & BRAIDWOOD Wes tin gh o u se Energy Sys tems W

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WESTINGHOUSE NOh-PROPRIETARY CLASS 3 WCAP-14824, Revision 1 Byron Unit 1 Heatup and Cooldown Limit Curves For Normal Operation and Surveillance Weld Metal Integration for Byron and Braidwood B P. A. Grendys April 1997 Work Performed Under Shop Order CPEP-139 Prepared by the Westinghouse Electric Corporation for the Commonwealth Edison Company Approved: 6 O C. H. Boyd, Manager 0~ N j 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 l

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l 1

F- 1 PREFACE i

i This report has been technically reviewed and verified by:

J T. J. Laubham ,

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• - L Byron Unit 1 Heatup and Cooldown Limit Curves - . April 1997 ,

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il l l TABLE OF CONTENTS iii LI ST O F FI GU R E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

LI ST OF TABLE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv 1

1 I NTR OD U CTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 2

2 FRACTURE TOUGHNESS PROPERTIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS . .

6 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE . . . . . . . . . . . . .

13 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES . . . .

21 6 R EFER ENC ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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 April 1997

' Byron Unit 1 Heatup and Cooldown Umit Curves

7 iii LIST OF FIGURES 1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100'F/hr) Applicable for tree First 12 EFPY (Without Margins for 15 Instrumentation Errors) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2 Byron Unit 1 Reactor Coolant System Cooldcwn i. imitations (Covidown Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for 16 Instrumentation Errors) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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 Instrumontation Errors: Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and the Reactor Vessel Beltline Region) . . . . . . . . . . . 17 4 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) AMable for the First 12 EFPY (Without Margins for tastrumentation Errors; Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and the Reactor Vessel Beltline Region) . . . . . . . . . . . 18 t

April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves

iv i

LIST OF TABLES 1 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1 7

Base M ate rials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1 Weld Material (Using Byron 1 & 2 Chemistry Test Results) . . . . . . . . . . . . . . . . . . 8 9

3 Byron Unit 1 Reactor Vessel Material Properties . . . . . . . . . . . . . . . . . . . . . . . . .

4 Calculation of Chemistry Factors Using Credible Byron Units 1 and 2 Surveillance Capsule Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5 Calculation of Adjusted Reference Temperatures (ART) at 12 EFPY for the Limiting Byron Unit 1 Reactor Vessel Material - Intermediate Shell Forging SP 5933 (based on credible surveillance capsu'e data) . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 1

Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T l 6

12 Locations for 12 EFPY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins for i 7

l Instrumentation Errors Includes 1) Vesset flange requirements of 180*F and 621 psig per 10CFR50. .............. 19 8 Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins for Instrumentation Errors includes 1) Vesset flange requirements of 180*l and 621 psig per 10CFR50. and 2) Pressure adjustment of 74 psig to account for pressure t neic.::e wween the wide-range pressure l

l transmitter and the limiting beltline region of the reactor vesset. . . . . . . . . . . . . . . . ...... 20 Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

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1 1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTuoy (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTuor of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, cstimating the radiation induced ARTuor, and adding a margin. The unirradiated RTuo,is '

designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or I 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, ARTuo, due to the radiation cxposure associated with that time period must be added to the unirradiated RTuo7 (IRTuoy).

The extent of the shift in RTuo,is enhanced by certain chemical elements (such as copper cnd nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials m. Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTuoy +

ARTuoy + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness  ;

I of the vessel at the beltline region measured from the clad / base metal interface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves.

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Byron Unit 1 HCup and Cooldown Limit Curve April 1997

1 2 i i

2 FRACTURE TOUGHNESS PROPERTIES i

l The fracture toughness properties of the ferritic material in the reactor coolant pressure )

boundary are determined in accordance with the NRC Regulatory Standard Review Plan:t 1, l j

The pre-irradiation fracture-toughness properties of the Byron Unit 1 reactor vessel are presented in Table 3. The post-irradiation fracture toughness properties of the reactor vessel beltline material were obtained directly from the Byron Unit 1 Reactor Vessel Radiation Surveillance Programts). Credible surveillance data is available for two capsules (Capsules U and X) for Byron Unit 1. This capsule data is used to calculate chemistry factors (See Table

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 programsH1 has been integrated pursuant to 10 CFR 50.61 in accordance with Regulatory Guide 1.99, Revision 2, Position 2.

In addition to the credible surveillance weld data from Byron Unit 1, credible surveillance weld data is available for two capsules (Capsules U and W) for Byron Unit 2. The chemistry factor values resulting from the weld metal integration for Byron Units 1 and 2 are presented in Section 4 of this report. See Tables 1 through 4.

A complete technical justification for the Byron Units 1 and 2 weld metal integration is presented in Appendix A of this report.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 i

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3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE I

RELATIONSHIPS i

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Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements'51 specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor l coolant pressure boundary of light water nuclear power reactors to provide adequate margins l of safety during any condition of normal operation, including anticipated operational f 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 *l, Vessels, contain the conservative methods of analysis.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K,, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference '

stress intensity factor, K , for the metal temperature at that time. K, is obtained from the f71 reference fracture toughness curve, defined in Appendix G of the ASME Code,Section XI .

The K, curve is given by the following equation: I i

to.o14s(7-nr,.1so)) (1)

K,=26.78 +1.233 + e where, {

K = reference stress intensity factor as a function of the metal temperature T and the {'

metal reference nil-ductility temperature RTuo, Therev.re, the governing equation for the heatup-cooldown er:clysis is defined in Appendix G of the ASME Code as follows:

(2)

C+ Kg K,<K,

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} K. = stress intensity factor caused by membrane (pressure) stress

! K, = stress intensity factor caused by the thermal gradients K = function of temperature relative to the RTuor of the material O= 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 April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves

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At any time during the heatup or cooldown transient, K,, is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RT,,, and the reference fracture toughness curve. The thermal stresses resulting from the temperatJre gradients through the vessel wall are calculated and then the corresponding 2

(thermal) stress intensity factors, K,, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

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 at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both

- steady state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

J The use of the composite curve in the cooldown analysis is necessary because control of the 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 vessel location is at a higher temperature than the

fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of K,, at the 1/4T location i for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K,, exceeds K,, the calculated allowable pressure during cooldown will be greeter than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.  ;

Three separate calculations are required to determine the limit curves for finite heatup rates.

As is done in the cooldown analysis, allowable pressure-temperature relationships are developed 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 internal pressure.

The metal temperature at the crack tip lags the coolant temperature; therefore, the K,, for the 1/4T crack during heatup is lower than the K,, for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

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K,, values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when -

f l the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to' ensure.

that at any coolant temperature the lower value of the allowable pressure calculated for i steady-state and finite'heatup rates is obtained.

The'second portion of the heatup analysis concems the calculation of the

'l pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the' l 4 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 j I

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.

8 Following the generation of pressure temperature curves' for both the steady state and finite -

heatup rate situations, the final limit curves are produced by constructing a composite curve -

. based on a point-by-point comparison of the steady-state and finite heatup rate data. At any

.l given temperature, the allowable pressure is taken to be the lesser of the three values taken j

from the curves under consideration. The use of the composite curve is necessary to set

. conservative heatup limitations because it is possible for conditions to exist wherein, over the j

. course of the heatup ramp, the controlling condition switches from the inside to the outside, j _ and the pressure limit must at all times be based on analysis of the most critical criterion.

I 10 CFR Part 50, Appendix G addresses the metal temperature of the closure head flange and j

i' 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 normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is

[ 621 psig for Byron Unit 1.

l 4

The limiting unirradiated RTwo, 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 l

i greater than 621 psig. This limit is shown in Figures 1 through 4 wherever applicable.

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April 1997

Byron Unit 1 Heatup and Cooldown Limit Curves i

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i 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

(3) i ART =InitialRTgARTgMargin j

Initial RTuor is the re'erence temperature for the unirradiated material as defined in paragraph W

NB-2331 of Section lit of the ASME Boiler and Pressure Vessel Code . If measured values of initial RTum for the material in question are not available, generic mean values for that f class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ARTuor is the mean value of the adjustment in reference temperature caused by irradiation ,

and should be calculated as follows:

A RT,- CF+ f(***'"" (4) d To calculate ART uor at any depth (e.g., at 1/4T or 3/4T), the following formula must first be  !

used to attenuate the fluence at the specific depth. j suano.*V e "'* (5) i(mwwuo=1 1

i where x inches (vessel beltline thickness is 8.5 inches!") is the depth into the vessel wall

' measured from the vessel clad / base metal interface. The resultant fluence is then placed in Equation 4 to calculate the ARTum 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 n/cm'. This fluence value was calculated from the surveillance Capsule X analysis presented in WCAP 13880M. i CF (*F) is the chemistry factor, obtained from the tables in Reference 1, using the average ]

i values of copper and nickel content as calculated in Tables 1 and 2 and reported in Table 3.

The chemistry factors were also calculated using the surveillance capsule data in Table 4.

The Ratio Procedure, as documented in paragraph (c)(2)(ii)(B) of 10 CFR Part 50.61, was used to adjust the measured values of 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.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

7 All materials '- the beltline region of Byron Unit 1 reactor vessel were considered in determining tne inniting material. Sample calculations to determine the ART values for the Weld Metal at 12 EFPY are shown in Table 5. The resulting ART values for all beltline regir materials at the 1/4T and 3/4T locations are summarized in Table 6, where it can be seen tha the limiting material is the Intermediate Shell Forging SP-5933 (based on credible surveillance capsule data). The 1/4T and 3/4T ART values for Intermediate Shell Forging SP-5933 (based on credible surveillance capsule data) will be used in the generation of heatup and cooldown curves applicable to 12 EFPY.

TABLE 1

. Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1 Base Materials intermediate M" Forging Lower Shell Forging Reference SP-5933 SP-5951 0.034 0.73 0.04 0.64 Byron Unit 1 HU/CD Limit Cunes 0.032 0.791 0.03 0.75 Letter Report FDRT/

0.05 0.73 SRPLO-009(94) 0.036 0.735 January 1994 Average 0.0364 0.747 0.04 0.64 Start tard Deviation 0.007 0.023 0 0 l

i I

I Byron Unit 1 Heatup and Cooldown Umit Curves April 1997

8

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TABLE 2 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 1 Weld Material (Using Byron 1 & 2 Chernistry Test Results)

Best-Estimate Reference pj! N_i 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&W Weld Qualification 0.053 0.62 B&W Weld Qualification 0.059 0.62 0.022 0.690 ---> 0.02 0.69 Sury. CF = 27 Byron 1 Surveillance Data See Below 0.023 0.712 ---> 0.02 0.71 Surv. CF = 27 Byron 2 Surveillance Data See Below Best-Estimate Chemistry: 0.057 0.614 --> 0.06 0.61 Best Est. CF = 82 Standard Deviation: 0.043 0.095 Byron 1 & 2 Ratio = 3.0 Surveillance Data Chemistry Results: Byron Unit 2 Byron Unit t Reference Q $

g WCAP-10398 O.03 0.65 Reference g WCAP 9517M 0.026 0.71 WCAP 12431A 0.024 0.740 WCAP-11651'84 0.023 0.67 0.024 0.786 0.022 0.665 0.022 0.704 0.021 0.714 0.020 0.681 0.741 0.021 0.706 0.021 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 0.020 0.706 0.020 0.678 0.021 0.677 0.020 0.695 0.023 0.677 0.010 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 14064"" 0.024 0.705 0.022 0.697 0.023 0.706 0.021 0.634 0.023 0.698 0.682 0.024 0.696 )

WCAP-138805 0.024 0.022 0.678 0.023 0.711 Q,Q25 Q,ZQ1 0.024 0.708 Average 0.022 0.690 0.024 0.716 0.024 0.715 0.024 0.707 0.024 0.720 0.024 0.717 0.024 0.711 0.024 0.706 0.024 0.707 Q.925 0.717  ;

Average 0.023 0.712 4 l

1 l

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 l

9 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 surveillance weld is identical to that used in the reactor vessel core region girth seam (WF-447). The weld wire is type Linde MnMoNi (Low Cu-P), heat number 442002, with a Linde 80 type flux, lot number 8064.

TABLE 3 Byron Unit 1 Reactor Vessel Material Properties Material Description Cu (%) Ni(%) Chemistry Initial Factor (*) RTuor MS Closure Head Flange -- 0.74 -- 60)

Vessel Flange -- 0.73 -- 10'c)

Intermediate Shell Forging 0.0364 0.747 23.8 40 SP 5933 Lower Shell Forging SP-5951 0.04 0.64 26.0 10 Circumferential Weld WF-336 0.06 0.61 82.0 -30 NOTES:

(a) Chemistry Factors are calculated from Cu and Ni values per Regulatory Guide 1.99, Revision 2.

(b) initial RTuo, values are measured values.

(c) Closure head and vessel flange initial RTuor values are used for considering flange requirementd51 for the heatup/cooldown curves.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

10 TABLE 4 Calculation of Chemistry Factors Using Credible Byron Units 1 and 2 Surveillance Capsule Data Capsule Capsule FPS Meas. FF* FF8 Material Fluence f A RTuoy A RTuo7 Inter. Shell U 3.72x10 O.727 0 0 0.529 i Forging SP-5933 X 1.39x10" 1.091 30 32.73 1.19 (Tangential)

U 3.72x10 O.727 0 0 0.529 Inter. Shell Forging SP 5933 30 32.73 1.19 X 1.39x10 1.091

,)

Sum: 65.46 3.44 Chemistry Factor'* = 65.46 + 3.44 = 19.0*F U 3.72x10 O.727 0 0 0.00 0.529 Byron 1 Weld Metal WF-336* 35 105 114.56 1.19 X 1.39x10 1.091 U 3 996x10 O.746 0 0 0.00 0.557 Byron 2 Weld Metal WF-4475 1.;!11x10* 90 94.77 1.110 W 1.053 30 Sum: 209.33 3.386 Chemistry Factor'* = 209.33 + 3.386 = 61.8'F NOTES:

(a) FF = Fluence Factor = f*8"4 (b) Byron Unit 1 ARTuor 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 j cross sections in 1994 and are documented in WCAP-14044"'l.

(c) Byron Unit 2 capsule fluence, FF, and ARTuor values were obtained from the surveillance Capsule W analysis (WCAP 14064'"h using the ENDF/B-V scattering cross sections.

(d) Chemistry Factor EtFF*ARTuor) + E(FF')

(e) Adjusted ARTuo, per Ratio Procedure of 10 CFR 50.61. Ratio = 3.0. See Table 2.

April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves

11 Margin is calculated as, M = 2 o ,' + 03'. The standard deviation for the initial RT um margin term, o,, is 0*F when the initial RTuor is a measured value, and 17'F when a generic value is available. The standard deviation for the ARTum margin term, o3 , is 17'F for the plate, and 8.5'F for the plate (half the value) when surveillance data is used. For welds, o3 is equal to 28'F when surveillance capsule is not used, and equal to 14*F when credible surveillance capsule data is used. a 3need not exceed 0.5 times the mean value of ART um.

TABLE 5 Calculation of Adjusted Reference Ternperatures (ART) at 12 EFPY for the Limiting Byron Unit 1 Reactor Vessel Material Intermediate Shell Forging SP-5933 (based on credible surveillance capsule data)

Parameter Values Operating Time 12 EFPY Material Intermediate Shell Forging SP-5933 Location 1/4T 3/4T Chemistry Factor, CF (*F) 19.0 19.0 Fluence, f (10 n/cm')

• 0.486 0.175 Fluence Factor, FF 0.799 0.538 15.2 10.2 ARTum = CF x ff (*F) initial RTum, I (*F) 40 40 Margin, M (*F) 15.2 10.2 Adjusted Rcfarence Temperature (ART), (*F) 70 60 per Regulatory Guide 1.99 Revision 2 NOTES:

(a) The Byron Unit 1 reactor vessel wall thickness is 8.5 inches at the beltline region.

(b) Fluence, f, is based upon fu (10 n/cm', E>1.0 MeV) = 0.810 at 12 EFPY.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

12 TABLE 6 Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T l Locations for 12 EFPY l Material 12 EFPY 1/4T ART 3/4T ART Intermediate Shell Forging 78 66 SP 5933 (RG Position 1)

using credible surveillance 70* 60" capsule data (RG Position 2)

l Lower Shell Forging SP-5951 52 38 (RG Position 1)

Circumferential Weld WF-336 92 58 (RG Position 1)

I using credible surveillance 47 31 capsule data (RG Position 2)

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.

Byton Unit 1 Heatup and Cooldown Limit Curves April 1997 i

13 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES

\

Pressure temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods"'I discussed in Section 3 and 4 of this report. This Cpproved methodology is also presented in WCAP 14040-NP-AD 1, 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 1 0 51 Figures 3 and 4 do include this pressure difference of 74 psig.

Figures 1 and 3 present the heatup curves without margins for instrumentation errors using a heatup rate of 100*F/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 rnust be met before the reactor is made critical.

The reactor must not be made critical until pressure-temperature combinations are to the right of the 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 goveming equation for the hydrostatic test is defined in Appendix G to Section XI of the ASME Code as follows:

(6) 1.5 K,< K, where, K is the stress intensity factor covered by membrane (pressure) stress, K = 26.78 + 1.233 e ! " * "* T * '8 H, T is the minimum permissible metal temperature, and RTuor is the metal reference nil-ductility temperature The criticality 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 April 1997 o

14 i

pressure temperature limits or core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature- ,

required for the inservice hydrostatic test, and at least 40*F higher than the minimum permissible temperature in the corresponding pressure temperature curve for heatup and cooldown calculated as described in Section 3 of this report. The minimum temperature for the inservice hydrostatic leak tests for the Byron Unit 1 reactor vessel at 12 EFPY is 203*F.

The vertical line drawn from these points on the pressure-temperature curve, intersecting a-  ;

curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core l 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 1 reactor vessel. The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 through 4 are presented in Tables 6 and

7.  ;

Additionally, Westinghouse Engineering has reviewed the minimum boltup temperature requirements for the Byron Unit 1 reactor pressure vessel. According to Paragraph G-2222 of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix G, the reactor vessel may be bolted up and pressurized to 20 percent of the initial hydrostatic test pressure ,

at the initial RTer of the material stressed by the boltup. Therefore, since the most limiting initial RTm Vdue is 60*F (closure head flange), the reactor vessel can be bolted up at this

, temperature.

s 4-4 s

?

i Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

. = , __ . .-. . - - - - -

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Byron Unit 1 Heatup and Cooldown Limit Curves -

April 1997

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Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

17

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Byron Unit 1 Heatup and Cooldown Lirnit Curves April 1997 a

18 i

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Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

19

' TABLE 7

. Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins  :

for Instrumsntation Errors ,

includes 1) Vessel flange requirements of 180*F and 621 psig per 10CFR50. t Cooldown Curves Heatup Curve Steady State 25F 50F 100F 100F Criticality. Limit Leak Test Limit '

T T P T P T P T P T P T P P 60 621 60 595 60 554 60 470 60 621 203 0 152 2000 65 621 65 610 65 570 65 489 65 621 203 0 203 2485 ,

70 621 70 587 70 509 70 621 203 0 70 621 75 621 75 605 75 531 75 621 203 0 75 621 80 621 80 554 80 621 203 671 ,

80 621 80 621 i 85 621 85 621 85 579 85 621 203 657 85 621 90 621 90 607 90 621 203 646 90 621 90 621 95 621 95 621 95 621 95 621 203 639 95 621 100 621 100 621 100 621 100 621 203 634 100 621 105 621 105 621 105 621 105 621 203 632 105 621 110 621 110 621 110 621 110 621 203 633 l 110 621 115 621 115 621 115 621 115 621 203 637 115 621 '

120 621 120 621 120 621 120 621 203 642 120 621 125 621 125 621 125 621 125 621 125 621 203 651 130 621 130 621 130 621 130 621 203 661 130 621 621 135 621 135 621 135 621 135 621 203 674 135 621 140 621 140 621 140 621 140 621 203 689 140 145 621 145 621 145 621 203 707 145 621 621 150 621 150 621 203 727 150 155 621 203 749 155 621 160 621 203 774 160 621 165 621 165 621 205 801 170 621 170 621 210 831 621 175 621 215 864 l 175 621 180 621 220 900 180 180 1483 180 900 225 938 185 1559 185 938 230 980 1640 190 980 235 1026 190 1728 195 1026 240 1075  ;

195  ;

200 1075 245 1128 200 1821 205 1128 250 1186 {

205- 1921 255 1247 l 210 2029 210 1186 i

215 2143 215 1247 260 1313 2266 220 1313 265 1385 1 220 '

225 1385 270 1461 225 2397 230 1461 275 1543 235 1543 280 1630 240 1630 285 1724 245 1724 290 1825 1 250 1825 295 1933 l 255 1933 300 2048 260 2048 305 2171 265 2171 310 2302 270 2302 315 2441 275 2441 i I

- (Configuration #9393315685880 for Cooldown, #2756858609292 for Heatup)

Byron Unit 1 Heatup a xi Cooldown Limit Curves April 1997

20 i

TABLE 8 Byron Unit 1 Heatup and Cooldown Data at 12 EFPY Without Margins for instrumentation Errors includes 1) Vessel flange requirements of 180'F and 62' psig per 10CFR50, and 2) Pressure adjustment of 74 psig to account for pressure difference between the wide-range pressure transmitter and the limiting bettline region of the reactor vessol.

Cooldown Curves Heatup Curve Steady State 25F 50F 100F 100F Criticality. Umit Leak Test Limit T T P T P T P T P T P T P P 547 60 521 60 480 60 396 60 547 203 0 182 2000 60 65 547 65 536 65 496 65 415 65 547 203 0 203 2485 70 547 70 513 70 435 70 547 203 0 70 547 75 547 75 531 75 457 75 547 203 0 75 547 80 547 80 547 80 480 80 547 203 597 80 547 85 547 85 547 85 505 85 547 203 583 85 547 90 547 90 547 90 533 90 547 203 572 90 547 95 547 95 547 95 547 95 547 203 565 95 547 100 547 100 547 100 547 100 547 203 560 100 547 105 547 105 547 105 547 105 547 203 558 105 547 110 547 110 547 110 547 110 547 203 559 110 547 115 547 115 547 115 547 115 547 203 563 115 547 120 547 120 547 120 547 120 547 203 568 120 547 547 125 5J.' 125 547 125 547 125 547 203 577 125 130 547 130 547 130 547 130 547 130 547 203 587 135 547 135 547 135 547 135 547 135 547 203 600 140 547 140 547 140 547 140 547 140 547 203 615 145 547 145 547 i45 547 145 547 203 633 150 547 150 547 150 547 203 653 155 547 155 547 203 675 160 547 160 547 203 700 165 547 165 547 205 727 170 547 170 547 210 757 175 547 175 547 215 790 180 547 180 547 220 826 180 1409 180 826 225 864 185 1485 185 864 230 906 190 1566 190 906 235 952 195 1654 195 952 240 1001 200 1747 200 1001 245 1054 205 1847 205 1054 250 1112 210 1955 210 1112 255 1173 215 2069 215 1173 260 1239 220 2192 220 1239 265 1311 225 2323 225 1311 270 1387 230 1387 275 1469 235 1469 260 1L56 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 (Configuration #9395568588093 for Cooldown, #9291115685880 for Heatup)

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

21  ;

l 6 REFERENCES .

1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel ,

. Materials", U.S. Nuclear Regulatory Commission, May,1988. l l

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.

3 WCAP 9517, " Commonwealth Edison Co. Byron Station Unit 1 Reacter ' vessel 1 Radiation Surveillance Program', J. A. Davidson, July 1979.  ;

i 4 WCAP-10398, " Commonwealth Edison Co. Byron Station Unit 2 Reactor Vessel ,

Radiation Surveillance Program", L. R. Singer, December 1983.

5 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", Federal Register, Volume 60, No. 243, dated December 19,1995.

6 1992 Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, Appendix G, " Vessels". j 7 .1989 ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix G,

" Fracture Toughness Criteria for Protection Against Failure".

8 1989 Section Ill, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph

- NB 2331, " Material for Vessels". .

i- 9 WCAP 13880, " Analysis of Capsule X from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", P. A. Peter, et al., January l 1994.

10 WCAP-14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation". E.

P. Lippincott, April 1994.

. 11 WCAP-14064, " Analysis of Capsule W from the Commonwealth 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.,

1 -. April 1975.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 -

i

. _ . , - E

--- - - -, - . - s ,.

22 13 WCAP-14040-NP-A, Revision 2, " Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D.

.' Andrachek, et al., January 1996.

14 Babcock & Wilcox drawing numbers 184557E, Rev. 2; 185266E, Rev. 2; 185297E, Rev.

2; 185328E, Rev. 2; " Reactor Vessel Longitudinal Section". .

15 Nuclear Safety Advisory Letter, NSAL 93-005A, " Cold Overpressure Mitigation System

.(COMS) Nonconservatism", L. R. Hardwick and H. A Sepp,3/10/93.

16 WCAP-14063, " Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation",

P. A. Peter, November 1994.

.17 WCAP-13881, " Evaluation of Pressurized Thermal Shock for Byron Unit 1", P. A. Peter, January 1994.

18 WCAP 14054, " Evaluation of Pressurized Thermal Shock for Byron Unit 2", P. A. Peter, August 1995.

19 WCAP-14242, " Evaluation of Pressurized Thermal Shock for Braidwood Unit 1", P. A.

Peter, March 1995.

20 WCAP-14229, " Evaluation of Pressurized Thermal Shock for Braidwood Unit 2", P. A.

Peter, March 1995.

21 WCAP-11651," Analysis of Capsule U From The Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko, et al.,

November 1987.

22 WCAP 12431, " Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program", E. Terek, et al., October 1989.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 a

o A-0 ,

e i

APPENDIX A WELD METAL INTEGRATION FOR BYRON UNITS 1 AND 2 s

P i

l 1

I l

4

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Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

~ - ..-.. . - - - . -. -.. . - . . - - - . - - - , .- -

A-1 :

i. INTRODUCTION:

l Westinghouse performed an evaluation to determine if the weld wire data of the Byron Units 1 7; 4 - 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  ;

4 program weld metal of each unit,

2. What vendor fabricated the welds and in what time frame, ,

- 3. What heat treatment did each surveillance program weld receive,  ;

4. Is the initial RTug of the welds the same or relatively close,

5. Is the initial upper shell energy of the welds the same or relatively close,

6. Is the geometry of the plants the same, ,

s

7. Is the type of fuel in all plants the same,

8. Are the fuel loading pattems in the plants similar (i.e., iow 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, -l 12; Can the criteria for credibility in 10 CFR Part 50.61 be met for each plant?

l 4

j l

l l

l Nyron Unit 1 Heatup and Cooldown Limit Curves April 1997

?

A  !

l

. EVALUATION:

1. What weld wire heat number, flux and flux lot numbers were used to fabricate the welds?

The surveillance program weld metal for each unit was fabricated with the following weld i wire and flux: l Byron 1: The weld metal is 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 weld (seam WF-336).

Byron 2: The weld metal is type Linde MnMoNi, heat number 442002, with a Linde 80 '

type flux, lot number 8064. This is the same heat number used in the limiting beltline weld (seam WF-447). l The Byron 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 and in what time frame ?

Byron .1: B&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 surveillance program weld metals were fabricated in the same 7 time frame and by the same vendor. Therefore, this information supports the integration of the surveillance program test results for these welds.

3. What heat treatment did each weld receive?

The surveillance program weld metals received the fcilowing 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 Byron 2: 1150

• 50'F for 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; fumace-cooled The post weld stress relief heat treatment given to the Byron 1 and 2 surveillance program welds was slightly different. However, based on engineering judgement, the slight ,

differences in temperature and time should not cause a significant difference in the material toughness properties.

4. Is the initial RTm of the 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), the initial RT, of the welds differ. However, the surveillance  ;

ma erials have performed similarly, and it is shift data that is used in the integration of -

1 Byron Unit 1 Heatup and Cooldown Limit Curves April 1997  ;

._ - i

A-3 i

data. As can be seen in Table 4 (page 10 of this report), the measured shifts in RT,are

. relatively the same. For example, the shift for the first capsules from Byron 1 and Byron 2 is 0*F. For the second capsules removed from Byron Units 1 and 2, the measured shifts are equal to 30*F and 35'F, respectively. These results are very close. Therefore, this l

~i information supports the 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?

Byron 1: 74 ft-lb Byron 2: 67 ft-lb l 1

The initial upper shelf energy values for the surveillance weld materials in the Byron j surveillance programs are very similar. Therefore, this infonnation supports the integration of the surveillance program test results for these welds.

6. Is the geometry 0: the plants the same?

Byron Units 1 ano 2 have a reactor vessel inner diameter of 173 inches, a reactor vessel beltline thickness af 8.5 inches (excluding the cladding). Both have a power rating of 3411 MWt and are Westinghouse 4 loop NSSS plants. Both vessels have nectron pads and the surveillance capsules are located at the same azimuthal angles.

7. Is the fuel design in allplants the same?

Byron 1 & 2 use 17X17 rod array fuel assemblies with the same fuel design, thus producing similar ' adiation effects at the surveillance capsules.

8. Are the fuelloading patterns in the plants similar (i.e. Iow leakage, etc.)?

Byron 1 & 2 use a low leakage loading pattem.

9. What is the projected 32 effective full power year surface fluence of each plant?

[

L 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.

Byron Unit 1 0* if 2F 39 49 1.290x10 1.947x10" 2.159x10 1.705x10" 1.939x10

Byron Unit 2 0* if 29 3r 4F

' 1.3f,3x10 1.979x10* 2.192x10 1.772x10 2.026x10

1 1

a-

~ April 1997

Symn Unit 1 Heatup and Cooldown Umit Curves

_y

A-4

10. What are the vesselinlet temperatures?

Byron 1: 558.4*F Byron 2: 558.4*F

11. What are the differences in the capsule lead factors of the plants?

Based ora 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 Location Lead Factor Capsule Location Lead Factor Capsule 3.85 U 58.5 3.96 U 58.5 X 238.5 3.79 W 121.5o 3.89 61.0 3.59 V 61.0o 3.64 V

241.0 3.59 Y 241.0 3.64 Y

238.5 3.89 W 121.5 3.79 X 301.5 3.79 Z 301.So 3.89 Z

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 vessel integrity of both units.

12. Can the criteria for credibility in to CFR Part 50.61 be met for each plant?

i Credibility will be evaluated for 1) all the surveillance capsule data (base metal & weld l metal) for Byron Unit 1, and 2) weld metal (only) for Byron Unit 2. The credibility determination will use the Byron Unit 2 weld metal data for the Byron 1 heatup/cooldown pressure temperature limit curves. Therefore, it must be determined to be credible.

Criterion 1: The materinIs in the survaillance capsules must be those which are controlling materials with regard to radiation embrittlement.

The following is a list of the beltline materiels contained in the Byron Units 1 and 2 surveillance programs:

. I April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves 1 _ _ . _

A-5 ,

Byron Unit 1: Intermediate shell forging SP-5933 Circumferential weld seam WF-336, heat number 442002, with a Linde 80 type flux, lot number 8873. (This is the esme heat number used in the

. limiting beltline weld.)

Byron Unit 2: Intermediate shell forging 49D329/49C297-1-1 Circumferential weld seam WF-447, heat number 4 ' 2002, with a Linde 80 type flux, lot number 8064. (This is the same heat r umber used in the limiting beltline weld.)

Based on the calculated RT,1, values presented in WCAP-13881 (Byron 1 PTS) and the ,

information provided in the Byron Unit 2 material selection documents, these materials are judged to be the most controlling with regard to radiation embrittlement for each unit.

Therefore, Criteria #1 is met for both units.

Criterion 2: Scatter in 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-9517 " Commonwealth Edison Co. Byron Station Unit 1 Reactor Vessel Radiation i Surveillance Program," dated July 1979 and WCAP-10398, " Commonwealth Edison Co. Byron Station Unit 2 Reactor Vessel Radiation Surveillance Program," dated December 1983. Plots of Charpy energy versus temperature for the irradiated conditicos are presented in the WCAP reports for Capsules U & X (Unit 1) and U & W (Unit 2).

Based on engineering judgement, the scatter in the data presented in these reports is small enough to determine the 30 ft lb temperature and the upper shelf energy of the Byron Units 1

& 2 surveillance weld metals unambiguously. Therefore, the Byron Units 1 & 2 surveillance materials meet this criteria.

l Criterion 3: When there are two or more sets of surveillance dets from one reactor, the scatter of ARTm 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 or more 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.

1 t

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

A-6 TABLE A-2*

Byron Units 1 & 2 Surveillance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 FFN Measured FFx FF' Material Capsule f*

(x) ARTuor 4 3Tuoy (x')

(y) (xy) l 3.72x10 O.727 0 0 0.529 Byron Unit 1 U Inter. Shell Forging 1 Mx10 1.091 30 32.73 1.190 SP-5933 (Axial) X fl.72x10 O.727 0 0 0.529 Byron Unit 1 U Inter. Shell Forging X t.39x10" 1.091 30 32.73 1.190 SP-5933 (Tangential) '

I",,, 3.636 60 65.46 3.44 U 3.72x10 O.727 0 0.00 0.529 Byron Unit 1 Weld Metal X 1.39x10 1.091 35 38.185 1.190 U 3.996x10 O.746 0 0.00 0.557 Byron Unit 2 Weld Metal W 1.211 x10 1.053 30 31.600 1.110 I",.i '"J 65 69.785 3.386 blOTES:

(a) f = Fluence (10" n/cm', E > 1.0 MeV)

'N (b) FF = Fluence Factor = f***

(c) Values of Iand ARTuoy for Byron 1 were taken from WCAP-14044 and WCAP-13880, respectively. The Byron Unit 2 values were taken from Table 3 of WCAP-14063.

i Syron Unit 1 Heatup and Cooldown Limit Curves April 1997 L --. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _

'A-7 ,

Per the 27I Edition of the CRC Standard Mathematical Tables (page 497), for a straight line fit by the method of least squares, the valuet h,and b, are obtained by solving the normal equations:

n b, + b, E x, = E y,

.Cnd b, Ex;+ b, Ex,8 = Ixy, These eqections can be re-written as follows:

n n r

[ y, - an + b[61 x, i=1 and n n n

[ X,y, = a[ X, + b[ X, 61 i=1 i=1 Bvron 1 & 2 Weld Metal:

. Based on the data provided in Table A 2 the equations become: f

1. 65.0 = 4a + 3.617b or a = 16.25 - 0.9043b and

2. 69.785 = 3.617a + 3.386b  ;

Thus, by substituting Eq.1 into Eq. 2, b = 95.71. Now, enter b (= 95.71) into Eq.1 and a = - .

70.30. Therefore, the equation of the straight line which provides the best fit in the sense of ,

least squares is:

Y' = 95.71 (X) - 70.30 The error in predicting a value Y corresponding to a given X value is: e = Y - Y' j Byron 1 Base Metal:

Based on the data provided in Table A 2 the equations become:

1. 60.0 = 4a + 3.636b or a = 15.0 - 0.909b and

2. 65.46 = 3.636a + 3.44b Thus, by substituting Eq.1 into Eq. 2, b = 80.89. Now, enter b (= 80.89) into Eq.1 and a = -

58.53. Therefore, the equation of the straight line which provides the best fit in the sense of least squares is:

i Y' = 80.89 (X) - 58.53

The error in predicting a value Y corresponding to a given X value is: e = Y - Y' Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 l

A-8 TABLE A-3 Best Fit Evaluation for Byron 1 & 2 Surveillance Materials Base Material ARTuo, Best Fit Scatter of j FF (30 ft-lb) ('F) A RT,, (*F) ARTu or W) l l

muummmmisemmun u mmuummmmmmmmuumummmmmmmmmmmmmum- l 0.727 0 -0.72 -0.00 Byron 1 & 2 Weld Metal -0.00 35.00 1.091 35 1

0.746 0 -0.00 0.00 1.053 30 -0.00 30.00 J l

0.727 0 0.28 -0.28 Byron Unit 1 Inter. Shell Forging SP-5933 (Axial) 29.72 0.28 1.091 30 Byron Unit 1 inter. Shell 0 0.28 -0.28 0.727 Forging SP-5933 1.091 30 29.72 0.28 (Tangential)

Weld Metal:

The scatter of ARTum values about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 28'F for weld metal. As shown above, the error is within 28 F of the best-fit line. Therefore, this criteria is met for the Byron Units 1 & 2 surveillance weld material.

Base Materigh The scatter of ARTum values about a best-fit line drawn, as described in Regulatory Position 2.1, should be less than 17 F for base metal. As shown above, the error is within 17'F of the best-fit line. Therefore, this criteria is met for the Byron Unit 1 surveillance base metal.

See the following scatter plots for the Byron Unit 1 base material and the Byron 1 and 2 weld metal.

i l

I

~

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 -

i l A-9 l

l 1

)

l

. . _ . . _ . . . . _ _ _ _ _ . _ . _ , _ _ _ _ - . - . - - _ _ _ . . . ~ - - - - --

Byron 1 Base Material 30 ,e

. 25

/ e Byron 1 SP 5933 (Axial) .

Measured

/

W 20 - / 3 Byron 1 SP-5933 (Amal)- Best-  ;

! / I8I i

, K 15 /  ! AByron 1 SP-5933 (Tangentia!) i j

/

Measured I

to .

X Byron 1 SP-5933 (Tangential) ; '

; Best-Fit 5 / \

/ .'

0 W - I O 0.5 1 1.5 Fluence (1019, E>1.0MeV)  ;

.- - -- - ..-- . -~~ -

I Byron 1 & 2 Weld Metal ,

35 g 30 ,5 25 ,/

/

• 20 i e Byron 1 & 2 Weld Metal- I f

h E 15

/  ; Measured l

/ !3 Byron 1 & 2 Weld Metal Best-l l

{

1 10

/

/ l j

i f

/

/ i 0

0 h 0.5 ,

1 1 5 l

-5 '  ;

Fluence (1019, E>1.0MeV)  !

1 I

(

)

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 I

A.10 Criterion 4: The Irradiation temperature of the Charpy specimens in the capsule must equal the vessel wall temperature at the cladding / base metal Interface within +/ 2S*F.

The Byron 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. The test capsules are in baskets attached to the neutron pad. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel well 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.

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 materint.

Byron Units 1 & 2 did not incorporate correlation monitor material in their surveillance program. Therefore, Criterion 5 is not applicable.

RESULTS & CONCLUSIONS:

Based on the evaluation performed above, it has been determined that there is sufficient data to support integrating the Byron Unit 1 weld metal surveillance data with Byron Unit 2 weld metal surveillance data.

i

- Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 L_-

A-11 j EFFECT OF WELD METAL INTEGRATION ON BYRON P-T LIMIT CURVES:

Plant Previous Previous New New Results 1/4T ART 3/4T ART 1/4T ART 3/4T ART 60'" --' ' '

Byre : 1 66.37*' 57.15*' 70'*

Crves at 8EFPY FDRT/SRPLO.

009(94)

Byron 2 43.5 33.2 92.6 75.8 Current curves / PTS Curves at evaluation are NOT conservative. Using 16 EFPY weld metal integration WCAP 14063 will be more restrictive.

Byron Unit 2 curves to be regenerated and i documented in i WCAP 14881 NOTES:

(a) Even after weld metalintegration, still forging-limited. Weld metal integration has no effect.

(b) Calculated at 8 EFPY.

(c) Calculated at 12 EFPY.

The new ART values for Byron Unit 2 are significantly larger. A reasonable applicability date cannot be determined. New curves are to be generated for Byron Unit 2. The results will be j documented in WCAP-14881, ' Byron Unit 2 Heatup and Cooldown Curves for Normal f Operation".

Byron Unit 1 He ,iup and Cooldown Limit Curves April 1997 a

A 12 l

EFFECT OF WELD METAL INTEGRATION ON BYRON PTS CALCULATIONS:

The weld metalintegration CF values were calculated in Section 4 of this report. Specifically, the following weld metal CF values were used to determine the RTyr, values-l RG Position 1 CF RG Position 2 CF Byron Units 1 and 2 82.0*F 81.8'F i i The vessel material data used in the latest PTS evaluation reports '7'81was used in this evaluation. (No new material property values were calculated.) However, for the Byron Units 1 and 2 RTris calculations at 48 EFPY, new fluence values were interpolated to 48 EFPY.

The vessel surface fluence results reported in Section 6.0 of the latest Byron Unit 1 Mand Byion Unit 2 t"1 surveillance capsule analysis reports were used.

TABLE A-4 RTp73 Values for Byron Unit 1 CF f* FF" RTer, M A RT,73 RT,7s Material (*F) (*F) ('F) (*F) (*F) 1 32 EFPY intermediate Shell Forging 23.8 2.159 1.209 40 28.8 28.8 97.6 f SP 5933 l 80.1 Using surv. capsule data M 19.1 2.159 1.209 40 17 23.1 l

Lower Shell Forging SP-5951 28 2.159 1.209 10 31.4 31.4 72.8 l

82.0 2.159 1.209 -30 56 99.1 125.1 l Weld Metal WF-338 l

Using sury. capsule data

• 61.8 2.159 1.209 -30 28 74.7 72.7 48 EFPY Intermediate Shell Forging 23.8 3.238 1.309 40 31.2 31.2 102.4 SP 5933 Using surv. capsule data

• 19.1 3.238 1.309 40 17 25.0 82.0 Lower SheII Forging SP-5951 26 3.238 1.309 10 34.0 34.0 78.0 l ,

Weld Metal WF 336 82.0 3.238 1.309 -30 56 '

107.3 133.3 Using surv. capsule data

• 61.8 3.238 1.309 30 28 80.9 78.9 Nw es:

(a) 2.159 x 10 n/cm' (E>1.0 MeV) for 32 EFPY from Byron 1 PTS report (WCAP 13881). The following calcu'ation

. to obtain the 48 EFPY fluence value:

I 2.159x10 + (2.159x10* 3 807x10)*(48 - 32 EFPY) = 3.238x10 n/cm' 32 - 5.64 EFPY (b) FF (Fluence factor) = f'"'N (c) Calculated using a CF based on surveillance capsule data per Regulatory Guide 1.99. Revision 2, Position 2.

Byron Unit 1 Heatup and Oooldown Umit Curves April 1997

A-13 TABLE A-5 RTers VALUES FOR BYRON UNIT 2 CF (*F) f') FF*) RT,citui M(*F) ARTns RTns Material (*F) (*F) (*F) 32 EFPY Lower Shell Forging 32.2 2.192 1.213 -20 34.0 39.1 53.1 MK 24-3 Using sury. capsule data") 19.8 2.192 1.213 -20 17 24.0 21.0 20.0 2.192 1.213 -20 24.3 24.3 28.6 Inter. Shell Forging MK 24-2 82.0 2.192 1.213 10 56 99.5 165.5 Cire. Weld Metal WF447 Using surv. capsule data") 61.8 2.192 1.213 10 28 75.0 113.0 48 EFPY 32.2 3.288 1.312 -20 34.0 42.2 56.2 Lower Shell Forging MK 24-3 Using surv. capsule data") 19.8 3.288 1.312 -20 17 26.0 23.0 20.0 3.288 1.312 20 26.2 26.2 32.4 Inter. Shell Forging MK 24-2 82.0 3.288 1.312 10 56 107.6 173.6 Cire. Weld Metal WF447 Using surv. capsule data") 61.8 3.288 1.312 10 28 81.1 119.1 NOTES (a) 2.192 x 10 n/cm' (E>1.0 MeV) for 32 EFPY from Byron 2 PTS report (WCAP-14054). The following calculation to obtain the 48 EFPY fluence value:

2.192x10 + (2.192x10" - 3.174x10l'(48 32 EFPY) = 3.288x10 n/cm' 32 4.634 EFPY (b) FF (Fluence factor) = f***'**

(c) Calculated using a CF based n surveillance capsule data per Regulatory Guide 1.99, Revision 2, Position 2.

l Byron Unit 1 Heatup and Cooldown Lirnit Curves April 1997

B-0 APPENDIX B WELD METAL INTEGRATION FOR BRAIDWOOD UNITS 1 AND 2 l

l l

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l r-l Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

B-1 INTRODUCTION:

Westinghouse performed an evaluation to determine if the weld wire data of the Braidwood Units 1 and 2 surveillance programs can be integrated. The evaluation was based on the following criteria: ,

1. What weld wire heat number, flux, and flux lot were used to fabricate the surveillance program weld metal of each unit, .

' 2. ~ What vendor fabricated the welds and in what time frame, ,

3. What heat treatment did each surveillance program weld receive, 1 4 Is the initial RTuor of the welds the same or relatively close,

5. Is the initial upper shelf energy of the welds the same or relatively close,

6. Is the geometry of the plants the same,

7. Is the type of fuel in all plants the same,

8. Are the fuel loading pattems in the plants similar (i.e., low leakage, etc.),

9. What is the projected 32 effective full power year surface fluence of each plant,

10. What vessel inlet temperatures do the plants operate at,

11. What are the differences in the capsule lead factors of the plants, l 1

12. Can the criteria for credibility in 10 CFR Part 50.61 be met for each plant?

i Byron Unit 1 Heatup and Cooldown Limit Curves . April 1997

=_ __= - - _ _ _ _ _ .- . _.. .. = _ -

B-2 EVALUATION:

1. What weld wire heat number, flux and flux lot numbers were used to fabricate the welds?

Braidwood 1: The weld metal is classification EF2N Low Cu, MnMoNi Heat number 442011, with a Linde grade 80 type flux, lot number 8061. This is the same heat number used in th9 limiting beltline weld (seam WF-562).

Braidwood 2: 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).

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 and in 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 integrction 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; furnace 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. Hcwever, 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 RTa of the welds the same or relatively close?

Braidwood 1: 40'F Braidwood 2: 40*F ,

The Braidwood Units 1 and 2 initial RT, values are identical. Therefore, this information supports the integration of the surveillance program test results for these welds.

Byron Unit 1 Heatup and Cooldown Umit Curves April 1997

B3

5. Is the initial upper shelf energy of the surveillance welds the same or relatively close?

Braidwood 1: 70 ft Ib j Braidwood 2: 71 ft-lb 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.. ]

6. Is the geometry of the plants the same?

All four plants have a reactor vessel inner diameter off 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 pa6s and the surveillance capsules are located at the same azimuthal angles. <

l 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.

8. Are the fuelloading pattems in the plants similar (i.e. Iow leakage, etc.)?

Braidwood 1 & 2 use a low leakage loading pattern.

9. What is the projected 32 effective full power year 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.

Braidwood Unit 1 0* 19 29 39 49 1.321x10 1.984x10 2.239x10 1.86Ex10 2.162x10*

Braidwood Unit 2 0* 19 29 3F 49 1.299x10 1.924x10 2.199x10 1.861x10 2.174x10

10. What vesselinlet temperatures do the plants operate?

Braidwood 1: 558.4'F Braidwood 2: - 558.4'F -

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.

Byron Unit 1 Heatup and Cooldown Limit Curves - April 1997

B-4 TABLE B-1 Surveillance Capsub Lead Factors for Braidwood Units 1 & 2 Braidwood Unit 1 Braidwood Unit 2 Capsule Location Lead Factor Capsule Location Lead Factor 58.5 4.03 U 58.5o 4.00 U

I 238.5 4.03 X 238.5 4.02 X

W 121.So 4.03 W 121.50 4.02 301.5 4.03 Z 301.So 4.02 Z i V 61.0 3.73 V 61.0 3.70 l 241.0 3.73 Y 241.0 3.70 Y

Based on the projected vessel surface fluence and lead factor values for Braidwood 1 & 2, the Braidwood 1 & 2 surveillance capsules will have approximately the same flux rates and irradiation temperatures. This supports the use of the surveillance weld data in both programs to evaluate the reactor vessel integrity of the Braidwood units.

12. Can the criteria for credibility in 10 CFR Part 50.61 be met for each plant?

Credibility will be evaluated for the Braidwood Units 1 and 2 weld metal (only) to show that Braidwood 1 & 2 can share weld metal data and determine an integrated weld metal chemistry factor.

Criterion 1: The materials in the surveillance capsules must he those which are controlling materials with regard to radiation embrittlement.

The following is a list of the beltline materials contained in the Braidwood Units 1 and 2 surveillance programs:

Breidwood Unit 1: Lower shell forging 49D867/49C813-1-1 Circumferential weld seam WF 562, heat number 442011, with a Linde grade 80 type flux, lot number 8061. (This is the same heat number used in the limiting beltline weld.)

Braidwood Unit 2: Lower shell forging 50D102/50C97-1-1 Circumferential weld seam WF-562, heat number 442011, with a Linde grade 80 type flux, lot number 8061. (This is the same heat number used in the limiting beltline weld.)

Based on the calculated RT,n values presented in WCAP-14242 (Braidwood 1 PTS) and the information provided in the Braidwood Unit 2 material selection documents, these materials ,

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997 o - _ _ _ _ _

B-5

]

l

)

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: Scatter in 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 Radiation Surveillance Program," dated February 1981 and WCAP-11188, " Commonwealth j 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.

Based on engineering judgement, the scatter in the data presented in these reports is small enough to determine the 30 ft lb temperature and the upper shelf energy of the Braidwood Units 1 & 2 surveillance weld metals unambiguously. Therefore, the Braidwood Units 1 & 2 surveillance materials meet this criteria.

Criterion 3: Where there are two or more sets of surveillance dets from one reactor, the scatter of ART,7 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 or more 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. l Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

B-6 l

l TABLE B-2#

Braidwood Units 1 & 2 Surveillance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 Capsule f(*) FF*) Measured FF x FF' ,

Material  !

(x) ART. A RT,,, (x*)

(y) (xy)

Braidwood 0.3814x10" 0.733 10 7.333 0.538 l U

Unit 1 1.144x10" 1.038 25 25.95 1.077 Weld Metal X 0.3933x10" 0.741 0 0.00 0.550 Braidwood U Unit 2 1.067 X 1.126x10" 1.033 20 20.66 Weld Metal I",, 3.545 55 53.943 3.232 Chemistry Factof = 53.943 + 3.232 = 16.7 NOTES:

(a) f = Fluence (10" n/cm', E > 1.0 MeV)

(b) FF = Fluence Factor = t' "' N (c) Values of f, FF, and ART, values were taken from Table 2 of WCAP 14243 (Braidwood Unit 1 P-T Limig) and WCAP-(4730 (Braidwood Unit 2 P-T Limits).

(d) CF = L(FF*RT ) + L(FF')

Per the 27E Edition of the CRC Standard Mathematical Tables (page 497), for a straight line fit by the method of least squares, the values b, and b, are obtained by solving the normal equations n bo + b, Ex, = Ey, and b E x, + b, E x,8 = E xy, These equations can be re-written as follows:

n n

[ y, - an + b[ x, let 41 and i n n n

[ x,y, = a[ x, + b[ x, 11 i=1 i=1 s

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

B-7 Braidwood 1 & 2 Weld Metal:

l Based on the data provided in Table B-2, the equations become:

55.0 = 4a + 3.545b or a = 13.75 - 0.8863b and l 1.)

2.) 53.943 = 3.545a + 3.232b Thus, by aubstituting Eq.1 into Eq. 2, b = 57.69. Now, enter b (= 57.69) into Eq.1 and a = -

37.38. Therefore, the equation of the straight line which provides the best fit in the sense of least squares is:

Y' = 57.69 (X) - 37.38 l The error in predicting a value Y corresponding to a given X value is: e = Y - Y' 1

I I TABLE B-3 l Best Fit Evaluation for Byron 1 & 2 and Braidwood 1 & 2 Weld Metal Base Material ART,, Best Fit Scatter of FF (30 ft lb) ('F) A RTum ('F) A RT,, ('F) summmmmmmmumsmummmmmmmmmmmmmmusummmmmmmmmmmu-0.733 10 -0.00 10.00 Braidwood 1 a 2 Weld Metal 1.038 25 0.00 25.00 0.741 0 -0.00 0.00 1.033 20 -0.00 20.00 Weld Metal:

The scatter of ARTuor values about a best-fit line drawn, as described in Regulatory l

Position 2.1, should be less than 28'F for weld metal. As shown above, the error is within 28'F of the best-fit line. Therefore, this criteria is met for the Braidwood Units 1 & 2 surveillance weld material.

See the following plot of ARTuo, versus fluence.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

.. i

B-8 Braidwood 1 & 2 Weld Metal 25 0

, 20 $

• i e Bradwood i & 2 Wekt Metal.

h15  ; Measured E ;e Braidwood 1 & 2 Weld Metal-

+ Best-Fit 10 _

5 0 0 0 ,0.2 0.4 0.6 0.8 1 1.2 Fluence (1019, E>1.0Mov)

Criterion 4: The Irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding @ase metal interface within +/-25*F.

The Braidwood 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. The test capsules are in baskets attached to the neutron pad. The location of the 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.

Criterion 5: The surveillance data for the correlation monitor materialin the '

capsule should fall within the scatter band of the data base for that material. +

Braidwood Units 1 & 2 did not incorporate correlation monitor material in their surveillance program. Therefore, Criterion 5 is not applicable.

RESULTS & CONCLUSIONS:

. Based on the evaluation performed above it has been determined that there is sufficient data to support integrating the Braidwood Unit 1 weld metal surveillance data with Braidwood Unit 2 weld metal surveillance data.

(

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

B-9 TABLE B-4 Calculation of Average Cu and Ni Weight Percent Values for the Braicwood Weld Material (Using Braidwood 1 & 2 Chernistry Test Results)

Best-Estimate Reference C.y Nj B&W Weld Qualification BAW 2261 0.020 0.63 B&W Weld Qualification 0.03 0.65 B&W Weld Qualification 0.04 0.67 Braidwood 1 Surv. Data See Below 0.032 0.671 --> 0.03 0.67 Surv. CF = 41 Braidwood 2 Sury. Data See Below 0.033 0.708 --> 0.03 0.71 Surv. CF = 41 Bost-Estimate Chemistry: 0.033 0.666 --> 0.03 0.67 Best Est. CF = 41 Standard Deviation: 0.005 0.029 Braidwood 1 & 2 Ratio = 1.0 Surveillance Chemistry Results:

Braldwood Unit 1 Braldwood Unit 2 Ni Reference Cu Ni Reference p_y 0.04* 0.67* WCAP 11188 0.040 0.64 WCAP-9807 WCAP 12685 0.035 0.666 WOAP-14228 0.033 0.724 0.033 0.666 0.034 0.711 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 l 0.031 0.683 0.032 0.743  ;

0.031 0.730 j 0.032 0.673 0.029 0.668 0.032 0.711 0.029 0.686 0.032 0.728 0.034 0.616 0.031 0.703 0.033 0.651 0.032 0.687 0.033 0.698 0.033 0.703 0.031 0.656 0.033 0.695 WCAP-14241 0.031 0.655 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 0.029 0.644 0.028 0.648 0.032 0.699 ,

0.027 0.644 0,034 0.765 {

l 0.034 0.668 0.031 0.673 {

0.033 0.656 0.034 0.724 l 0.035 0.747 i

0.036 0.658 '

0.036 0.671 0.033 0.711 9.,f92 0.031 0.688 9.11Q Average 0.032 0.671 0.035 0.750 9All Q_681 Average 0.033 0.708 i

i Not used in Average calculation; reported for completeness. The same value appears in the material test reports and the surveillance program report.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

B-10 EFFECT OF WELD METAL INTEGRATION ON BRAIDWOOD P-T LIMIT CURVES:

Previous New New Result Plant Pmvious 1/4T ART 3/4T ART 1/4T ART 3/4T ART 76.6 65.4 69.7 60.6 Current curves / PTS  ;

Braidwood 1 l evaluation are Curves at  !

conservative.

16 EFPY New Applicability Date:

WCAP 14243 27.9 EFPY 55.7 69.5 60.4 Current curves / PTS l Braidwood 2 62.6 evaluation are NOT f Curves at conservative. Using 16 EFPY weld metal integration will be more restrictive.

WCAP-14230 New Applicability Date:

7.4 EFPY After the Braidwood Units 1 and 2 surveillance weld metal is integrated, the following calculations show the new applicability dates of the heatup/cooldown pressure-temperature limit curves.

BRAIDWOOD UNIT 1:

Weld Metal calculations based on a 1/4T ART = 76.6*F:

(The following data is from Braidwood Unit 1 heatup/cooldown curve report, WCAP-14243) i Per Regulatory Guide (RG) 1.99, Revision 2: ART = 1 + M + (CF

• FF)

Using the " Previous" ART values and initial RT , this equation was used to back-calculate the fluence factor (FF) and the vessel surface fluence value. This fluence value was then used to determine a new applicability date (in terms of EFPY) for the current pressure-temperature limit curves.

For Braidwood Units 1 and 2, the 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.

April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves

B i 8

The Margin term is calculated as, M = 2 (o,' + 03 ). The standard deviation for the initial RTuo, margin term (c) is 0*F when the initial RTuor is a measured value (as is the case for the Byron units). Additionally, the term o, need not exceed 0.5 times the mean value of A RTum.

Therefore, when the ARTum value is multiplied by 0.5 and plugged into the above equation, the effect is 2 * (ARTum /2), which is the ARTum (or CF

• FF).

ART = 1 + (CF

• FF) + (CF

• FF) l 76.6*F = 40*F + (16.7

• 1/4T FF)*F + (16.7

• 1/4T FF)*F ==> 1/4T FF = 1.0958  :

8 1.0958 = 1/4T f** *" * "; 0 ==> 1/4T f = 1.4124 x 10 n/cm 8

1.4124 x 10 = f

• e* ' " ' S ==> f = 2.352 x 10 n/cm This fluence value will occur after 32 EFPY, per Table 6-15 of WCAP-14241. The following calculation will determine the applicability date in terms of EFPY.

Fluence at X EFPY = Fluence at 32 EFPY + (X - 32 EFPY)

• Fluence /EFPY 2.352 x 10 = 2.239 x 10 + (X - 32 EFPY) * (2.239 x 10 - 1.120 x 10'$

32 - 16 EFPY X = 33.6 EFPY i

1 l

l I

b I

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

_: j

B-12 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)

ART = 1 + M + (CF

• FF) 65.4*F = 40*F + (16.7

• 3/4T FF)*F + (16.7

• 3'4T FF)*F ==> 3/4T FF = 0.76047 0.76047 = 3/4T f 5#* W m o ==> 3/4T f = 0.4221 x 10" n/cm' O.4221 x 10 = f

• eM2"'#*** ==> f = 1.9493 x 10 n/cm' 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.

Fluence at X EFPY = Fluence at 16 EFPY + (X - 16 EFPY)

• Fluence /EFPY 1.9493 x 10 = 1.120 x 10 + (X - 16 EFPY) * (2.239 x 10- 1.120 x 10'S 32 - 16 EFPY X = 27.9 EFPY Therefore, after the weld metal integrr. tion for Braidwood Units 1 and 2 is implemented, the Braidwood Unit 1 heatup/cooldown curves presented in WCAP-14243 will be applicable to 27.9 EFPY.

April 1997_ ~

Byron Unit 1 Heatup and Cooldown Limit Curves a

B-13 BRAIDWOOD UNIT 2 Weld Metal calcu.d!:m based on a 1/4T ART = 62.6*F:

(The following data is from Braidwood Unit 2 heatup/cooldown curve report, WCAP-14230.)

ART = 1 + M + (CF

• FF) 62.6*F = 40*F + (16.7

• 1/4T FF)*F + (16.7

• 1/4T FF)'F ==> 1/4T FF = 0.6766 0.6766 = 1/4T f*28 '"' * * * ==> 1/4T f = 3.075 x 10" n/cm' 2

3.075 x 10 = f

• eta 24 es

• 02o ==> f = 5.120 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.

Fluence at X EFPY = Fluence at 4.215 EFPY + (X - 4.215 EFPY)

• Fluence /EFPY 5.120 x 10" = 2.896 x 10" + (X - 4.215 EFPY) * (1.100 x 10"- 2.896 x 10'*)

16 - 4.215 EFPY X = 7.4 EFPY Weld Metal calculations based on a 3/4T ART = 55.7'F:

ART = 1 + M + (CF

• FF) 55.7'F = 40*F + (16.7

• 3/4T FF)*F + (16.7

• 3/4T FF)'F ==> 3/4T FF = 0.47006 0.47006 = 3/4T f*2' * * * ==> 3/4T f = 0.1292 x 10" n/cm' O.1292 x 10" = f

• e'*2 '* * "M ==> f = 5.966 x 10" n/cm' This fluence value will occur between 4.215 and 16 EFPY, per Table 615 of WCAP-14228.

The following calculation will determine the applicability date in terms of EFPY.

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 10")

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 EFPY.

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

B 14 EFFECT OF WELD METAL INTEGRATION ON BRAIDWOOD PTS CALCULATIONS:

The weld metal integration CF values were calculated in Section 4 of this report. Specifically, the following weld metal CF values were used to determine the RTets values:

RG Position 1 CF RG Position 2 CF 41.0'F 16.7'F Braidwood Units 1 and 2 The vessel material data used in the latest PTS evaluation reports"* was used in this evaluation. (No new material property values were calculated.)

TABLE B-4 RTyr, Values for Braidwood Unit 1 FF) RT yu M ARTers RTpts CF f

('F) (*F) ('F) ('F)

Material (*F)

~

32 EFPY 1.218 30 34 37.77 41.8 Inter. Shell Forging 31.0 2.239 2.239 1.218 20 31.68 31.68 43.4 Lower Shell Forging 26.0 2.239 1.218 20 17 22.90 19.8 using S/C data *' 18.8 1.218 40 49.95 49.95 139.9 Weld Metal WF-562 41.0 2.239 2.239 1.218 40 20.34 20.34 80.7 using S/C data *3 16.7 48 EFPY 3.358 1.317 -30 34 40.83 44.8 Inter. Shell Forging 31.0 3.358 1.317 -20 34 34.25 48.3 Lower Shell Forging 26.0 3.358 1.317 -20 17 24.76 21.8 using S/C data *) 18.8 41.0 3.358 1.317 40 54.00 54.00 148.0 Weld Metal WF-562 3.358 1.317 40 21.99 21.99 84.0 using S/C data *' 16.7 NOTES, (a) FF (Fluence factor) = f**"'

(b) Calculated using a CF based on surveillance capsule data per Regulatory Guide 1.99, Revision 2, Position 2.

1 April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves l L . . _ _ _ _ _ _ _ _ _ _ _ _

B-15 TABLE B-5 RTers Values for Braidwood Unit 2 f FF" M(*F) A RT,7s RTris CF ('F) RTworcui Material ('F) (*F) (*F) 32 EFPY 20.0 2.199 1.214 30 24.28 24.28 18.6 Upper Shell Forging 2.199 1.214 -30 34 44.92 48.9 Lower Shell forging 37.0 13.3 2.199 1.214 -30 16.15 16.15 2.3 using S/C data")

41.0 2.199 1.214 40 49.77 49.77 139.5 Weld Metal WF 562 16.7 2.199 1.214 40 20.27 20.27 80.5 using S/C data *)

48 EFPY 20.0 3.298 1.313 -30 26.26 26.26 22.5 Upper Shell Forging 37.0 3.298 1.313 -30 34 48.58 52.6 Lower Shell Forging 13 3 3.298 1.313 -30 17 17.46 4.5 using S/C data *'

41.0 3.298 1.313 40 53.83 53.83 147.7 Weld Metal WF-562 16.7 3.298 1.313 40 21.93 21.93 83.9 using S/C data *'

NOTES, (a) FF (Fluence factor) = f"''* *

(b) Calculated using a CF based n surveillance capsule data per Regulatory Guide 1.99. Revision 2, Position 2.

l I

i Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

C-0 I

APPENDlX C q BYRON /BRAIDWOOD FLUENCE METHODOLOGY JUSTIFICATION AND TIME-DEPENDENT CAPSULE FLUENCE VALUES 1

1 I

l 1

i

\

l l l

l April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves

}

l C .1 I

l, 1 Fluence Methodoloav Justification lJ The fast neutron exposure methodology documented in WCAP-14040-NP A, " Methodology L

Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" is consisunt with the requirements of Draft Regulatory Guide DG--

1053, " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" and makes use of neutron transport cross-sections derived from the ENDF/B-VI data base. The exposure evaluations documented in WCAPs 13880,14064,14241, and 14228 for the Byron Units 1 & 2 and Braidwood Units 1 & 2 pressure vessels were completed prior to

- the release of the ENDF/B VI based Light Water Reactor neutron transport cross-section library, Consequently the neutron transport calculations performed as an integral part of these evaluations were based on the currently available ENDF/B-IV based cross section library. In all respects other than the ENDF/B-VI vs ENDF/B-IV cross-section issue, the methodology 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.

It is planned that neutron fluence evaluations for the Byron Units 1 & 2 and Braidwood Units 1

& 2, praecure vessels will be updated to incorporate the use of ENDF/B-VI cross-section {

libraries at the time of the next scMaled surveillance apsule withdrawal for each of the units. Based on recent experience in updating vessel fluence evaluations to the ENDF/B VI i

enethodology for reactors of similar design to Byron and Braidwood it is anticipated that best estimate neutron fluence values will be impacted by less than 7% compared to those over those previously reported. Likewise, application of the ENDF/B-VI methodology to the re-evi%ation of neutron dosimetry from previously withdrawn surveillance capsules will change reported capsule exposures by an amount less than the uncertainties quoted with the prior dosimetry analyses. 1 In addition to the methodology upgrade discussed in the preceding paragraph, the fluence updates for Byron Units 1 & 2 and Braidwood Units 1 & 2 will also include an evaluation of low leakage fuel management instituted at all four units. A qualitative examination of the loading pattoms used 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 not effect of methods upgrades and low leakage fuel management on i

Byron Unit 1 Heatup and Cooldown Limit Curves April 1997

C2

. projectM vsssel fluence is, therefore, anticipated to be very small and may result in an overall rede.; tion in fluence relative to that reported in WCAPs 13880,14064,14241, and 14228.

Based on the relatively small changes that are anticipated from updating the neutron fluence evaluations from those reported in WCAPs 13880,14064,14241, and 14228 to the approved methodology described in WCAP-14040-NP-A, including the impact of low leakage fuel management, coupled with the low sensitivity to irradiation damage exhibited by the materials comprising the Byron Units 1 & 2 and Braidwood Units 1 & 2 reactor pressure vessels, the use of the previously documented fluence values is justified until the update to the ENDF/B-VI based methodology is completed for each unit.

2 ~- Time Deoendent Surveil!ance Caosule Fluences '

Based on the documentation provided in WCAPs 13880,14064,14241, and 14228, it is noted that the last surveillance car,sule withdrawal for Byron Units 1 & 2 and Braidwood Units 1 & 2 was at 5.64,4.63,4.23, and 4.21 effective full power years, respectively. Projection of fluence

. levels at the surveillance capsule locations for times beyond those withdrawal dates are needed in order to establish appropriate withdrawal schedules for the remaining capsules comprising the Reactor Vessel Surveillance Program for each of the units. These Best Estimate projections are provided in Tables C-1 through C-4 for Byron Units 1 & 2 and Braidwood Units 1 & 2, respectively. These projections are based on the assumption that the best estimate neutron flux averaged over the total irradiation time for each unit would remain applicable for the remainder of plant lifetime.

Ap?1997 1 Byron Unit 1 Heatup and Cooldown Limit Curves-L _ _ _ _ _ _ _ _ _ .

C-3 i

3 TABLE C 1 l BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECT!ONS AT SURVEILLANCE CAPSULE LOCATIONS - BYRON UNIT 1 Irradiation Fluence [n/cm'J Lead Factor Time t 31.5 Caos 29.0 Caos 31.5 Caos 29.0 Caos $

[EFPY) 5.64- 1.443e+19 1.365e+19 3.79 3.58 8.00 2.046e+19 1.935e+19 3.79 3.58 10.00 2.558e+19 2.419e+1G 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 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 l l.

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 i

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April 1997 Byron Unit 1 Heatup and Cooldown Limit Curves i

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C-4 TABLE C-2 q BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS

- AT SURVEILLANCE CAPSULE LOCATIONS - BYRON UNIT 2 i

i irradiation Fluence [n/cm'] Lead Factor Time IEFPY1 31.5 Caos 29.0 Caos 31.5 Caos 29.0 Caps f 4.63 1.235e+19 1.154e+19 3.89 3.64 >

6.00 1.598e+19 1.494e+19 3.89 3.64 8.00 2.131e+19 1.992e+19 3.89 3.64 .

10.00 2.664e+19 2.491e+19 3.89 3.64 i 12.00 3.197e+19 2.989e+19 3.89 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 .

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 i

30.00 7.992e+19 7.472e+19 3.89 3.64 32.00 8.525e+19 7.970e+19 3.89 3.64  ;

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Byron Unit .1 Heaiup and Cooldown Limit Curves April 1997  :

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L I

C-5 1

TABLE C-3 BEST ESTIMATE FAST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS AT SURVEILLANCE CAPSULE LOCATIONS - BRAIDWOOD UNIT 1 Irradiation Fluence [n/cm'] Lead Factor Time -

IEFPY1 31/iCaos 29.0 Caos 31.5 Caos 29.0 Caos 4.23 1.193e+19 1.105e+19 4.02 3.73 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 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 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 9.014e419 8.348e+19 4.02 3.73 Byron Unit 1 Heatup and Cooldown Lhnit Curves . April 1997

C-6 TABLE C-4 BEST ESTIMATE F ST NEUTRON FLUENCE (E > 1.0 MeV) PROJECTIONS AT SURVEILLANCE CAPSULE LOCATIONS - BRA!DWOOD UNIT 2 Irradiation Fluence [n/cm'] Lead Fac*or Time 31.5 Caos 29.0 Caos 31.5 Caos 29.0 Caos

[EFPY1 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 t 18.00 4.968e+19 4.577e+19 4.02 3.70 i 20.00 5.520e+19 5.085e+19 4.02 3.70 22.00 6.072e+19 5.594e+19 4.02 3.70 24.00 6.625e+19 6.102e+19 4.02 3.70 t 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 l

4 i

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i Byron Unit 1 Heatup and Cooldown Umit Curves April 1997 :

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