ML20207H801

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Rev 0 to WCAP-15124, Byron Unit 1 Heatup & Cooldown Limit Curves for Normal Operation
ML20207H801
Person / Time
Site: Byron Constellation icon.png
Issue date: 11/30/1998
From: Christopher Boyd, Laubham T, Trombola D
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20207H753 List:
References
WCAP-15124, WCAP-15124-R, WCAP-15124-R00, NUDOCS 9907210213
Download: ML20207H801 (53)
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Westinghouse Non-Proprietary Class 3 WC AP- 15124 8evision 0 Byron Unit 1 Heatup and z

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Cooldown Limit Curves for Normal Operation . .

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9 Wes tin g h o u s e Energy . Systems W 9907210213 990712 PDR ADOCK 05000454 ',

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Westinghouse Non-Proprietary Class 3 f

i WCAP- 15124 . l Revasion 0 e

i Byron L nit 1 Heatup and 1 Cooldown Limit Curves for .

Normal Operation .

W Westinghouse Energy Systems ,

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9907210213 990712

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PDR ADOCK 05000454 )

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WESTINCHOUSE NON-PROPRIETARY CLASS 3 I

. WCAP-15124 l

Byron Unit 1 Heatup and Cooldown Limit Curves for l Normal Operation i

i T. J. Laubham November 1998  ;

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i Work Performed Under Shop Order C8QP-139 Prepared by the Westinghouse Electric Company i for the Commonwealth Edison Company l

Approved:

C. H. Boyd, ManSer Equipment & Materials Technology

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Approved:

D. M. Trombola, Manager

Mechanical Systems Integration i

l Westinghouse Electric Comy,4ny Energy Systems P.O. Box 355 Pittsburgh, PA 15230 4355

@1998 Westmghouse Electric Company All Rights Reserved

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TABLE OF CONTENTS l LIST OF TABLES . .. .... . .. ... . . . . . . . ... .. . . . . .. .. . . . . . . . . .iv '

LIST OF FIG URES . . . . . .. . . . . . . . . ... ... . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi P RE FA C E . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . .. . . . . . . . . . . . . . . . ... .vii EXEC UTIVE S UMMARY . .. . . . .... .... .. ... ... .... .... ..... . ... . . .. .... . ... . . . . . . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . vi 1

INTRODUCTION ... ...... ................. .. . .. . .......... .. ... .. . . . . . ..... . ................. . .. 1-1 2-PURPOSE................................................................................................................2-1 3-CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS... . . ... 3 1 1

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4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ... . . . . . . . . . . . .41 5

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES.. . . ......5-1 1

I 6- REFERENCES . . .. . .. .. ............. . ..... . . . . . . . . . . . . . . . . . . 6-1 1

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f LIST OFTABLES S -

Tr.ble 4-1 Summary of the Peak Pressu.e Vessel Neutron Fluence Values at 16 EFPY used for the Calculation of ART Values (n/cm', E > 1.0 MeV) . 4-4 l

l Table 4 2 Calculated Integrated Neutron Exposure of the Byron Unit 1 Surveillance Capsules i Tested to Date. . 4-4 Table 4 3 Measured 30 ft-Ib Transition Temperature Shifts of the Beltline Materials

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j Contained in the Surveillane.c Program . .4-5 '

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l Table 4-4 Calculation of Best Estimate Cu and Ni Weight Percent for the ByTon Urut i ,

Forging Material. 4-6 Table 4-5 Calculation of the Average Cu and Ni Weigat Percent for the Byron Urut i Surveillance Weld Matenal Only (Heat # 442002) . .4-6 Table 4-6 Calculation of Best Estimate Cu and Ni Weight Percent Values for the ByTon Units I & 2 Weld Material (Using Byron I & 2 Chemistry Test Results). .4-7 j Table 4-7 Reactor Vessel Beltline Material Unirradiated Toughness Properties . .4-8 Table 4-8 Calculation of Chemistry Factors using Byron Unit i Surveillance Capsule Data . . . .

.4-9 Table 4 9 Summary of the Byron Unit i Reactor Vessel Beltline Matenal Chenustry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1. 4-11 r

. e Table 4-10 Calculation of the I/4T and 3/4T Fluence Factors Values used for the Generation of the 16 EFPY Heatup and Cooldown Curves.. 4.I2 Table 4-11 Calculation of the ART Values for the 1/4T Location @ 16 EFPY.. 4 13 j Table 412 Calculation of the ART Valaes for the 3/4T location @ 16 EFPY 4 14 -

Table 4-13 Surnmary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location .4-15 Table 5-1 Byron Urut i Heatup Data at 16 EFPY Using 1989 App. G Methodologs (Without Margins for Instrumentatiori Errors) . 5-5 l

Table 5 2 Byron Urut i Cooldown Data at 16 EFPY Using 1989 App. G Methodology (Without Margins for Instrurnentauon Errors). 5-7 Revision 0

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v Table 5 Byron Unit i Heatup Data at 16 EFPY Using 1996 App. G Methodology (Without Margins for Instrumentation Errors). . . . . . .5 10 Table 5-4 Byron Unit I Cooldown Data at 16 EFPY Using 1996 App. G Methodology (Without Margins for lastrumentation Errors).. ..... . .. . . . . . . . . . .5 12 1

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LIST OF FIGURES Figure 5 l Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100'F/hr)

Applicable to 16 EFPY Using 1989 App. G Methodology (Without Margins for Instrumentation Errors).. .

.5-3 Figure 5 2 Byron Unit i Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100*F/hr) Applicable to 16 EFPY Using 1989 App. G .

Methodology (Without Margms for instrumentation Errors).. . 5-4 Figure 5-3 i Byron Unit i Reactor Coolant System He? tup Limitations (Heatup Rate of 100*F/hr) -

Applicable to 16 EFPY Using 1996 App. O Methodology t (Without Margins for Instrumentation Errors).. .5 8 Figure 5-4 Byron Unit i Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100*F/hr) Applicable to 16 EFPY Using 1996 App. G Methodology (Without Margms for Instrumentation Errors).. .59  ;

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c vit PREFACE -

This report has been technically reviewed and verified by;

- Reviewer: Ed Terek M-Revision 0 l

rn um EXECUTIVE

SUMMARY

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The purpose of this report is to generate pressure temperature limit curves for Byron Unit I for normal I operation at 16 EFPY using the methodology from WCAP 14040-NP-A which encompasses the requirements of the 1989 ASME Boiler and Pressure Vessel Code,Section XI Appendix G. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART). This port also presents pressure-temperature limit curves that follow the requirements of the 1996 Addenda to Appendix G for calculating the stress intensity factors. The 1/4T and 3/4T values are summarized in Table 4-13 and were calculated using the intennediate shell forging SP-5933 (i.e. The limiting beltline region material). The pressure-temperature limit curves were generated for a heatup rate of 100*F/hr and cooldown rates of 0,25,50 and 100'F/hr. These curves can be found in Figures 5-1 through 5-4.

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'1_ ilNTRODUCTION '

Heatup and cooldown limit curves are calculated using the adjusted RT:cn (reference nil ductility

- temperature) corresponding to the limiting beltime region material of the reactor vessel. The adjusted

< o limiting material in the core region of the reactor vessel is detennined by using the RTm f the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTu, and adding a margin. The unirradiated RTmyr 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 ofimpact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.

RTw increases as the material is exposed to fast-neutron radiation. Therefore to fmd the most limiting RTm t aany time period in the reactor's life, ARTm due to the radiation exposure associated with that time period must be added to the unirradiated RTa(IRTm). The extent of the shift in RTm is enh=ared by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a' method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials"Ut Regulatory Guide 1.99, Revision 2, is used for the calculation ofAdjusted Reference Temperature (ART) values (IRTmyr + ARTm + margms for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad / base metal interface. .The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves for normal operation.

NOTE: For the reactor vessel radiation surveillance program, Babcock and Wilcox Co. supplied

"/dg.ause with sections of SA508 Class 2 forging material used in the core region of the Byron Station unit No. I reactor pressure vessel (Specifically from forging SP-5933). Also supplied was the non-copper coated weld wire using the automatic sub-arc welding process (Weld wire heat # 442002 Linde 80 flux, lot j number 8873, which is identical to that used in the actual fabrication of the intermediate to lower shell girth weld of the pressure vessel).

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I Introduction Revision 0 i

2-1 2 PURPOSE The. Commonwealth Edison Company contracted Westinghouse to analyze surveillance capsule W from the Byrcn Unit I reactor vessel. As a part of this analysis Westinghouse generated new heatup and cooldown curves for 16 EFPY using the methodology from both the 1989 and 1996 ASME B&P Vessel Code,Section XI, Appendix G. The heatup and cooldown curves were generated without margins for instrumentation errors. The curves include a hydrostatic leak test limit curve from 2485 to 2000 psig and pressure temperature limits for the vessel flange regions per the requirements of 10 CFR Part 50, Appendir. GI23 The purpose of this report is to present the calculations and the development of the Commonwealth Edison Company Byron Unit I heatup and cooldown curves for 16 EFPY. This report documents the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2V3, for all the beltline materials and the development of the heatup and cooldown pressure-temperature limit curves for normal operation.

Per the request of the Commonwealth Edison Company, the surveillance weld data from the Byron Unit 1, Byron Unit 2, Braidwood Unit I and Braidwood Unit 2 surveillance programs has been integrated.

Note that Byron Unit 2 surveillance weld is identical to the surveillance weld (Heat No. 442002) at Byron Unit 1, and the Braidwood Units' surveillance weld is identical to the nozzle shell forging to intermediate shell forging girth weld (Heat No. 442011) at Byron Unit 1. Per WCAP 15123tti, all the surveillance weld data has been determined to be credible, while the Byron Unit I surveillance forging material data

. was determined to be non-credible.

Purpose Revision 0  !

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3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS-Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements"D3 specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary oflight water nuclear power ' reactors to provide adequate margins of safety during any condition j

. 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", Appendix GDI, contains the conservative methods of analysis.

The ASME approach for calculatmg the allowable limit curves for various heatup and cooldown rates specifies that the total stress atensity factor, Ki, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor. Ku, for the metal temperature at that time. Ku is obtamed from the reference fra:ture toughness curve, defmed in Appendix G of the ASME Code,Section XI. The Ku curve is given by the following equation: i Ku = 26.78 + 1.233

  • eruiwr-ar-+im (;)

where, Ku = reference stress intensity factor as a function of the metal temperature T and the metal reference nil <iuctahty L.w.. hire RTun Therefore, the governing equation for the heatup-cooidewn analysis is defmed in Appendix G of the ASME Code as follows:

l C *K m+Kn < Ku (2) where, .

Ki,,, = stress intensity factor caused by membrane (pressure) stress Ku = stress intensity factor caused by the thermal gradients

- Ku = function of temperature relative to the RTun of the material L

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 I

Criteria For Allowable Pressure-Temperature Relationships Revision 0 t

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At any time during the heatup or cooldown transient, Ki. 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 RTmn, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors. Ka, 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 ofAppendix 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 fmite cooldown rate situations.

From these relations, composite limit curves are constructed for each cooldown rate ofinterest.

The use of the composite curve in the cooldown analysis is necessary I,ecause 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. Durmg cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, ofcourse, is not tme 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 Ki. at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in Ki. exceeds Ka, the calculated allowable pressure during cooldown will be greater than the steady-state value.

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

Three separate calculations are required to determme the limit curves for finite heatup rates. As is done in I the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assummg the presence of a 1/4T defect at the inside of j

the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses j produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; j therefore, the Ki, for the 1/4T crack during heatup is lower than the K . for the 1/4T crack during i 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 Kr. 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 the 1/4T flaw is considered. i Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure e=1=l=W for steady-state and finite heatup rates is obtamed 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. U the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses wiuch are tensile in nature and eerefore tend to reinforce any pressure stresses Criteria For Allowable Pressure-Temperature Relationships Revision 0

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3-3 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

' increasmg 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 fmite heatup rate situations, the fmal limit curves are produced by constructing a composite curve based on a point by-point comparison of the steady-state and fmite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken 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 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 RTer by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psig), which is 621 psig for the Byron Unit I reactor vessel.

The limiting unirradiated RTer of 60'F (Table 4-7) occurs in the closure head flange of the Byron Unit I reactor vessel, so the muumum allowable temperature of this region is 180'F at pressures greater than 621 psig.

1996 Addenda to ASME Section XI. Annandiv G Methadalonvl "I Appendix G was recently revised to incorporate the most recent clastic solutions for Ki due to pressure and radial thermal gradients. The new solutions are based on fmite element analyses for inside surface flaws I

performed at Oak Ridge National Laboratories and sponsored by the NRC, and work published for outside surface flaws. These solutions provide results that are very similar to those obtained by using solutions previously developed by Raju and Newman I'%

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This revision now provides consistent computational methods for pressure and thermal Ki for thermal gradients through the vessel wall at any time during the transient. Consistent with the original version of Appendix G, no contribution for crack face pressure is included in the Ki due to pressure, and claddmg effects are neglected.

Using these most recent clastic solutions in the low temperature region will provide some relief to restrictions associated with reactor operation at relatively low temperatures. Although the reliefis re!sively small in terms of absolute allowable pressure, the benefits are substantial because even a small increase in the allowable pressure can be a signi6 cant percentage increase in the operatmg window at relatively low temperatures. Implementing this revision results in an economic and potential safety benefit (less likelihood oflifting LTOP relieving devices) with no reduction in vessel integrity; i.e. as an input to LTOP set points, the improvement in steady state maximum allowable pressure for Byron Unit lat 60'F is 40 psig.

Cnteria For Allowable Pressure-Temperature Relationships Revision 0 l.

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The following revisions were made to ASME Section XI, Appendix G:

G-2214.1 Membrane Tension:

K w = M x (pR, / t) (3) where, M. for an inside surface flaw is given by:

, M. =  ? 85 for E < 2, M. =

0.926E for 2:; E s 3.464, M. ' = 3.21 for E > 3.464 sumlarly, M. for an outside surface flaw is given by:

M. = 1.77 for E < 2, M. = 0.893 E for 2s E s 3.464, -

M. = 3.09 for E > 3.464  !

and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

J G-2214.3 RadialThermal Gradient:

The maximum Ki produced by radial thermal gradient forthe postulated inside surface defect of G-2120 is Ku = 0.953x10-' x CR'x t'# where CR'is the cooldown rate in *F/hr., or for a postulated outside surface defect, Kn = 0.753x10 x HU x t 24 , where HU is the heatup rate in 'F/hr.

The through-wall temperature difference assocated with the maxunum thermal Ki can be determined from Fig. G-2214-1. The ts.wi. hire at any radial distance from the vessel surface can be determined from Fig.

G'-2214-2 for the maximum thermal Kr .

(a) The maximum thermal Ki relationslup and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in'G-2214.3(a)(1) and (2)..

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(b)- Alternatively, the Ki for radial thermal gradient can be calculated for any thermal stress distribution and at any W time during cooldown for a %-thickness inside surface defect using the relationship  ;

i Criteria For Allowable Pmsure-Temperature Relationships Revision 0

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1 3-5 f L = (l.0359Co + 0.6322Ci + 0.4753C: + 0.3855C3)

  • M H)

.or sunitarly, K 7 during heatup for a %-thickness outside surface defect using the relationslup:

L = (1.043Co + 0.630Ci + 0.481C + 0.401C3)

  • M (5) where the coefficients Co, Ci, C2and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

a(x) = Co + Ci(x / a) + C2(x / a)2 + C (x / a)' (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth l

Note, that equations 3,4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation wh~lalogy. Therefore, the P-T curve methodology is -kaap from that described in WCAP-14040 'l ISection 2.6 (equations 2.6.2-4 r.nd 2.6.3-1) with the exceptions just described above. Per Reference 13, the NRC has resiewed and accepted this eh~ialogy for Byron Unit 2 in January of 1998.

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1 4- CALCULATION OF ADJUSTED REFERENCE TEMPERATURE I

From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each matenal in the l l

beltline region is given by the following expression-ART - InitialRTwr - A RTar - Margin (7)

Initial RTw is the reference temperature for the unirradiated matenal as defmed m paragraph NB-2331 of ,

Section 111 of the ASME Boiler and Pressure Vessel Codel If measured values ofinitial RTm for the material in question are not available, generic mean values for that class of matenal may be used if there

are sufficient test results to establish a mean and standard deviation for the class.  !

l ARTm is the mean value of the adjustment in reference temperature caused by irradiation and is l calculated as follows: l A RTar = CF

  • f""" " gg; To calculate ARTm 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.

/% = f,,,

  • c* (9) i where x inches (vessel inner radius and beltline thickness is 86.625 inches and 8.5 inches respectively)" I is the depth into the vessel' wall measured from the vessel clad / base metal interface. The resukant fluence is then placed in Equation 4 to calculate the ARTm at the specific depth.

The Westinghouse Radiation Engineering and Analysis group evaluated the vessel fluence projections and P

the results are presented in Section 6 ofWCAP-15123 1. %e evaluation used the ENDF/B-VI scattering cross-section data set. His is consistent with the methods presented in WCAP-14040-NP-A,

" Methodology Used to Develop Cold Overpressure Mitigating System Serpoints and RCS Heatup and Cooldown Limit Curves"Hl. Table 4-1, herein, contain the calculated vessel surface fluence values along with the Regulatory Guide 1.99, Revision 2,1/4T and 3/4T calculated fluences used to calculate the ART values for all beltline materials in the Byron Unit I reactor vessel. Additionally, the calculated surveillance capsule fluence values are prescnted in Table 4-2.

Ratio Procedure and Temocrature Adiustment:

The ratio procedure, as documented in Regulatory Guide 1.99, Revision 2 Position 2.1, was used, where applicable, to adjust the measured values of ARTm of the weld materials for differences in copper / nickel

content. This Mjustment is perfermed by multiplying the ARTm by the ratio of the vessel chemistry 5 factor to the surveillance material chemistry factor.

Calculation of Adjusted Reference Temperature Revision 0

4-2 From NRC Industry Meetings on November 12,1997 and February 12.13 of 1998, procedural guidelines were presented to adjust the ARTm for temperature difference when using surveillance from one vessel applied to another vessel. The following guidance was presented at these industry meetmgs:

Irradiation temperature and fluence (or fluence factor) are first order environmental variables in assessing irradition damage... To account for differences in temperature between surveillance i specimens and vessel, an adjustment to the data must be performed. Studies have shown that for temperatures near 550'F, a l'F decrease in irradiation temperature will result in approximately a j l'Fincreasein ARTa. ,

For capsules with irradiation temperature of Tw, and a plant with an irradiation temperature of Tw, an adjustment to normalize ARTa. to T, is made as follows:

Temp. Adjusted ARTm = ARTa, + 1.0*( Tw - Ty) (10)

"Ihe irradiation temperatures from Byron and Braidwood Units 1 & 2 are presented in WCAP-14824, Revision 2. The average irradation temperature from each of the four Units and operating cycles in question is $53'F. Therefore, no ; .r..iore adjustment is required.

Chemistry F=cear:

The chemistry factor is obtamed from the tables in Regulatory Guide 1.99, Revision 2 using the best estimate average copper and nickel content as reported in Tables 4-4,4-6 and 4-7. 'Ihe chemistry factors were also calculated using Position 2.1 from the Regulatory Guide 1.99, Revision 2 using all available surveillance data. Position 2.1 chemistry factors are calculated in Table 4-8.

Calculation of Adjusted Reference Temperature Revision 0

1 43 Ekolanation of Marcin Term:

When there are "two or more credible surveillance data sets'" available for Byron Unit 1. Regulatory )

Guide 1.99 Rev. 2 (RGl.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 from RG1.99R2 is as follows:

M = 2da; + a?

Standard Deviation for Initial RTm Margin Term, ei I

If the initial RTm alues v are measured values, which they are in the case of Byron Unit 1, then o ris equal f to O'F. On the other hand, if the initial RTer values were not measured, then a generic value of 17'F (base metal and weld metal) would have been required to be used for on .

l Standard Deviation for ARTm Margin Term, a3 Per RGl.99R2 Position 1.1, the values of e6 are referred to as "28'F for welds and 17'F for base metal, except that e4 need not exceed 0.50 times the mean value of ARTer." The mean value of ARTwr is defmed in RGl.99R2 by Equation 2 and defmed herein by Equation 8.

Per RGl.99R2 Position 2.1, when there is credible surveillance data, o4 si taken to be the lesser of %

ARTer or 14'F (28'F/2) for welds, or 8.5'F (17'F/2) for base metal. Where ARTer gam a is defmed herein by Equation 8.

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Summary of the Margin Term ]

l Since og is taken to be zero when a heat-specific measured value ofinitial RTer are available (as they are in this case), the total margin term, based on Equation 4 of RGl.99R2, will be as follows:

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= Position 1.1: Lesser of ARTer or 56*F for Welds lesser of ARTm or 34*F for Base Metal

. Position 2.1: Lesser of ARTer or 28'F for Welds Lesser of ARTer or 17'F for Base Metal i

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TABLE 4-1 Summary of the Peak Pressure Vessel Neutron Fluence Values at 16 EFPY used fo'r the Calculation of ART Values (n/cm2 , E > 1.0 MeV)

Azimuth Surface %T %T l l

I Nozzle Shell Forging 123J218 2.96 x 10 l.78 x 10 6.41 x 10

Intermediate Shell Forging SP-5933 9.85 x 10 5.91 x 10 2.13 x 10 )

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Lower Shell Forging SP-5951 9.85 x 10 5.91 x 10" 2.13 x 10 l I

intermediate to Lower Shell Forging Circ. 9.85 x 10 5.91 x 10" 2.13 x 10

{

Weld Seam WF-336 (Heat 442002) {

l Nozzle Shell to Intermediate Shell Forging 2.96 x 10 l.78 x 10" 6.41 x 10" {

Cire. Weld Seam WF-501 (Heat 442011)

Note: All remaining vessel materials are below 1 x 10" n/cm 2, E > 1.0 MeV 1

TABLE 4-2 Calculated Integrated Neutron Exposure of the Byron Unit 1 Surveillance Capsules Tested to Date Capsule Fluence U 4.04 x 1018 n/cm2, (E > 1.0 MeV)

X 1.57 x 1019 n/cm2, (E > 1.0 meV)

W 2.43 x 1019 n/cm2,(E > 1.0 MeV)

Calculation of Adjusted Reference Temperature Revision 0

45 Contained in Table 4 3 is a summary of the Measured 30 ft Ib transition temperature shifts of the beltline materials. These measured shift values were obtained usmg CVGRAPH Version 4.11 *. which is a hyperbolic tangent curve-fitting programf j I

TABLE 4 3 Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program Material Capsule Measured 30 ft-lb Transition Temperature Shift" Intermediate Shell Forging SP-5933 U 28.55'F I

X' 9.82'F (Tangential Orientation) W 49.2'F l

Intermediate Shell Forging SP-5933 U 18.52*F

}

X 53.03*F (Axial Orientation) W 29.34*F

_ Surveillance Program U 5.61*F .

i Weld Metal X- 40.ll'F W 51.34'F -

Heat Affected Zone U -60.2*F X 13.45'F W 15.23*F Notes:

(a) From capsule W analysis resultsm, Calculation of Adjusted Reference Temperature Revision 0

4-6 Table 4-4 contains the calculation of the best estimate weight percent copper and nickel for the Byron Unit 1 base materials in the beltline region. Table 4-5 contains the calculation of the best estimate weight percent copper and nickel for the Byron Unit I surveillance weld material, while Table 4-6 presents the overall best estimate average for that heat of weld. Table 4-7 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RTer of the beltline materials and vessel flanges. The weight percent values of Cu and Ni given in Table 4-7 were used to generate the calculated chemistry factor (CF) values based on Tables 1 and 2 of Regulatory Guide 1.99, Revision 2. and presented in Table 4-9.

Table 4-8 provides the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Position 2.1, which are also summarized in Table 4-9.

TABLE 4-4N Calculation of the Best Estimate Cu and Ni Weight Percent for the Byron Unit 1 Forging Materials Intermediate Shell Forging SP-5933 Lower Shell Forging SP-5951 Reference Cu % Ni % Cu % Ni %

Ref. S) 0.0364) 0.747 O.04 0.64 Charpy AL-31*) 0.036 0.70 --- ---

Charpy AT-36*) 0.035 0.76 - --- ---

q Best Estimate Average 0.04 0.74 0.04 0.64 Note:

(a) This is the average of 5 data points.

(b) Charpy Specimens From Capsule W of ByTon Unit I (Ref. 7).

(c) The best estimste average was rounded per ASTM E29, using the " Rounding Method" TABLE 4-5 Calculation of the Average Cu and Ni Weight Percent for the Byron Unit 1 Surveillance Weld Material Only (Heat # 442002)

Reference Weight % Copper Weight % Nickel Ref. 54 ') 0.022 0.69 Charpy AW-36*) 0.026 0.72 Charpy AW-40*) 0.027 0.72 Surveillance Weld Average 0.022 0.69 Note:

(a) This is the average of 21 data points.

(b) Charpy Specimens From Capsule W of Byton Unit 1 (Ref. 7).

Calculation of Adjusted Reference Temperature ,

Revision 0

4-7

)

TABLE 4-6 ,

Calculation of Best Estimate Cu and Ni Weight Percent Values for the Byron Units 1 & 2 Weld Material (Using Byron I & 2 Chemistry Test Results)  !

1 Chemistry Type Reference Weight % Copper Weight % Nickel B&W WQ: BAW-2261 Ref. 5 0.024 0.70 Ref. 5 0.031 0.46 B&W WQ: BAW-2261 B&W WQ: BAW 2261 Ref. 5 0 03 0.72 B&W WQ: BAW-2261 Ref. 5 0.068 0.48 B&W WQ: BAW 2261 Ref. 5 0.053 0.62 B&W WQ: BAW-2261 Ref. 5 0.059 0.62 B&W WQ (from NDIT No. Ref.5 0.029 0.65 BYR97 346, Rev. 0)

Round Robin Data Ave. on Weld Ref. 5 0.038 0.658 WF-336 (from NDIT No. BRW-DIT-97-391, Rev 0)

Byron 1 Surveillance Data Ave." Table 4 5 0.022 0.69 Byron 2 Surveillance Ave." Ref. 5 0.023 0.712 i

BESTESTIMATE AVERAGE'S *P - ' O.04 O.63

  • NOTES:

(a) The weld material in the Byron Unit I surveillance program was made of the same wire and flux as the reactor vessel inter, to lower shell girth seam weld. (Weld seam WF-336. Wire Heat # 442002, Flux Type Linde 80, Flux Lot # 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, tot number 8064.

(c) The best estimate chemistry values were obtained using the average of averages" approach. In addnion the best estimate average was rounded per ASTM E29, using the " Rounding Method" (d) This average Cu and Ni excluded two data points (Cu = 0.114, Ni = 0.54 and Cu = 0.148, Ni = 0.60), see Ref.15.

Calculation of Adjusted Reference Temperature Revision 0

l 4-8 l

l TABLE 4-7 Reactor Vessel Beltline Material Unirradiated Toughness Properties Material Description Cu(%) Ni(%) Initial RTmn'"  ;

Closure Head Flange 124K358VA1 --- 0.74 60 Vessel Flange 123J219 val --- 0.73 10 Nozzle Shell Forging 123J218 N 0.05 0.72 30 j l

Intermediate Shell Forging SP-5933 0.04 0.74 40 {

Lower Shell Forging SP-5951 0.04 0.64 10 Intermediate to Lower Shell Forging Cire. 0.04 0.63 -30 Weld Seam WF-336 (Heat # 442002)

Nuzzle Shell to Intermediate Shell Forging 0.03 0.67 10 )

Circ. Weld Seam WF-501 (Heat # 442011)*

Byron Unit 1 Surveillance Program 0.02 0.69 ---

Weld Metal (Heat # 442002) ,

l Byron Unit 2 Surveillance Program 0.02 0.71 ---

Weld Metal (Heat # 442002)

Braidwood Units 1 & 2 Surveillance Program 0.03 0.67, 0.71 ---

Weld Metals (Heat # 442011) l l

Natst l

(a) TheinitialRTan values for the plates and welds are based on measured data per reference 5 and 9.

(b) Best Estimate Cu% / Ni% and initial RTun Per Referenec 5 and/or 9.

Calculation of Adjusted Reference Temperature Revision 0

4-9 TABLE 4-8 Calculation of Chemistry Factors using Byron Unit i Surveillance Capsule Data Material Capsule Capsule f" FF"' ARTxm"' FFMRTuor FF l

U 0.404 0.748 28.55 21.36 0.560 Intermediate Shell Forging SP-5933 X 1.57 1.124 9.82 11.04 1.263 (Tangential) W 2.43 1.239 49.20 60.96 1.535 U 0.404 0.748 18.52 13.85 0.560 Intermediate Shell Forging SP-5933 X 1.57 1.124 53.03 59.61 1.263 (Axial) W 2.43 1.239 29.34 36.35 1.535 SUM: 203.17 6.716 CFsr.sm = Z(Fl

  • RTwot) + I( FF ) = (203.17) + (6.716) = 30.3'F Byron Unit I Surv. Weld U 0.404 0.748 11.22 8.39 0.560 (5.61)'d' Material X 1.57 1.124 80.22 90.17 1.263 (40.11)'d' (Heat # 442002) W 2.43 1.239 102.68 127.22 1.535 (51.34)

id Byron Unit 2 Sury. Weld U 0.405 0.749 O) 0 0.561 Material W l.27 1.067 60.0 64.02 1.138 (30.0)

(Heat # 442002) SUM: 289.80 5.057 CFs., w.e4 u2ao2 = I(FF

  • RTum) + I( FF ) = (289.80) + (5.057) = 57.3'F Notes:

(a) Byron Unit I and 2 capsule fluences were updated as a part of the capsule W dosimetry analysis results (Ref. 7), (x 10 n/cm 2, E > 1.0 MeV).

(b) FF = fluence factor A fa2s.oi wo ,

(c) ARTworvalues are the measured 30 ft-lb shift values taken from Ref. 7.

(d) The Byron I & 2 surveillance weld metal ARTwot values have been adjusted by a ratio factor of 2.00.

No temperature adjustment are required.

I Calculation of Adjusted Reference Temperature Revision 0

r 4-10 l l

l TABLE 4 Contmued Calculation of Chemistry Factors using Byron Unit i Surveillance Capsule Data Material Capsule Capsule (* FFN ARTw1 FFMRTer FF Weld Heat 442011, WF-501 U 0.3814 0.733 10 7.3 0.538 Using Braidwood i Sury. X 1.144 1.038 25 26.0 1.077 Data Weld Heat 442011, WF 501 U 0.3933 0.741 0 0 0.550 Using Braidwood 2 X 1.126 1.033 20 20.7 1.067 Sury. Data SUM: 54.0 3.232 CFse w.wu2on = I(FF

  • RTm7) + E( FF ) = (54.0) + (3.232) = 16.7T" I

Notes.

(a) Braidwood Units 1 & 2 fluences were taken from WCAP-14824 Res. 2 (Ref. 5)

(x 10" n/cm 2, E > 1.0 MeV).

(b) FF = fluence factor = fa23.ei w o ,

(c) v ARTer alues are the measured 30 ft-lb shift values taken from Appendix B of Ref. 5. j (d) The Braidwood I & 2 surveillance weld metal ARTer values do not require a ratio factor or temperature j adjustment. '

(c) Per Reference 14, Comed reported to the NRC a chemistry factor of 17.0. The difference is a result of rounding and is negligible when used in the calculation of adjusted reference temperature.

l l

l 5 A Calculation of Adjusted Reference Temperature Revision 0

d 4-11 TABLE 4-9 Summary of the Byron Unit i Reactor Vessel Beltline Matenal Chemistry Factors Based on Regulatory Guide 1.99. Revision 2. Position 1.1 and Position 2.1 Material Chemistry Factor Position 1.1 Position 2.1 Nozzle Shell Forging 123J218 31.0*F -

Intermediate Shell Forging SP-5933 26.0 F 30.3*F Lower Shell Forging SP-5951 26.0'F -

I 54.0 F 57.3*F Intennediate to Lower Shell Forging Cire.

Weld Seam WF-336 (Heat 442002)

Nozzle Shell to Interme& ate Shell Forging 41.0 F 16.7'F Cire. Weld Seam WF-501 (Heat 442011)

Byron Unit 1 & 2 Surveillance Weld Metal 27.0 F -

Braidwood Unit 1 & 2 Surveillance 41.0'F -

Weld Metal i

l l

Calculation of Adjusted Refenece Temperature Revision 0 :

4-12 Contained in Table 4-10 is the summary of the fluence factors (FF) used in the calculation of ad.iusted reference temperatures for the Byron Unit I reactor vessel beltline materials for 16 EFPY.

TABLE 4-10 Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 16 EPFY Heatup/Cooldown Curves 1/4 T F 1/4T FF 3/4T F 3/4 T FF Azimuth -

(a/cm', E >1.0 MeV)

(m/ca'. E > 1.0 MeV)

Nozzle Shell Forging 123J218 1.78 x 10" 0.542 6.41 x 10" 0.334 Intermediate Shell Forging SP-5933 5.91 x 10" 0.853 2.13 x 10" 0.585 Lower Shell Forging SP-5951 5.91 x 10" 0.853 2.13 x 10" 0.585 Intonih to Lower Shell Forging 5.91 x 10" 0.853 2.13 x 10" 0.585 Cire. Weld Seam WF-336 (Heat 442002)

Nozzle ShelltoIntermedate Shell 1.78 x 10" - 0.542 6.41 x 10" 0.334 Forging Cire. Weld Seam WF-501 (Heat 442011)

Contamed in Tables 4-11 and 4-12 are the calculations of the ART values used for the generation of the 16 EFPY heatup and cooldown curves.

Calculation of Adjusted Reference Temperature Revision 0

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4-15 The intermediate shell forging 5P-5933 is the linuting beltline materials for all heatup and cooldown curves to be generated. Contained in Table 4-13 is a summary of the limiting ARTS to be used in the generation of the Byron Unit I reactor vessel heatup and cooldown curves.

TABLE 4-13 Summary ofAdjusted Reference Temperature (ART) at i/4T and 3/4T Location for 16 EFPY l

Material 16 EFPY I/4 TART 3/4T ART l l

Intermediate Shell Forging SP-5933 84 70 1

, - Using Surveillance DataN 100 N 92N 1

Lower Shell Ferging SP-5951 54 40 Circumferential Weld WF-336 o2 33

- Using Credible Surveillance Data 47 32 Circumferential Weld WF-501 54 37

- Using Credible Surveillance Data 28 21 form Braidwood I and 2 Nozzle Shell Forging 123J218 64 51 NOTES:

(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99. Revision 2, Position 2 along with a full margin since it was determined that this data was not credible and the Table chemistry factor was non conservative.

(b) These ART values were used to generate the Byron Unit I heatup and cooldown curves.

Calculation of Adjusted Reference Temperature Revision 0

c 5-1 l

5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT 1 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 discussed in Section 3 and 4 of this report. This approved methodology is also presented in

- WCAP-14040-NP-AH3, dated January 1996.

Figures 5-1 and 5-3 present the heatup curves (without margins for possible instrumentation errors) for a heatup rate of 100'F/hr using the 1989 Appendix G methodology and the_1996 Appendix G methodology, )

respectively. The curves are applicable for 16 EFPY for the Byron Unit I reactor vessel. Additionally, )

Figures 5-2 and 5-4 present the cooldown curves (without margms of for possible instrumentation errors) for cooldown rates of 0,25,50 and 100*F/hr using the 1989 Appendix G methodology and the 1996

~ Appendix G methodology, respectively. These curves are also applicable for 16 EFPY for the Byron Unit I reactor vessel. Allowsble combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 through 5-4. This is in addition to other criteria which must be met before the reactor is made critical, as discussed in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-1 and 5-3 (for the specific heatup rate being utilized). The straight-line portion of the criticahty limit is at the minimum pennissible 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 defmed in Appendix G to Section XI of the ASME Codel81as follows:

)

J 1.5Kw < Ku (11) where, Kw is the stress intensity factor covered by membrane (pressure) stress, i Ku= 26.78 + 1.233 l e ** ""' '*l, T is the minimum permissible metal temperature, and  ;

RTer is the metal reference nil ductility temperature i L i i

The criticality limit curve specifies pressure-temperature limits for core operation to proside additional margin during actual power production as specified in Reference 2. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the nummum temperature required for the inservice hydrostatic test, and at least 40'F higher than the muumum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3 of this report. The muumum temperature for the inservice hydrostatic leak test for the Byron Unit I reactor vessel at 16 EFPY is 233*F at 2485 psig using the 1989 App. G Methodology and 225'F at 2485 psig using the 1996 App. G Methodology. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

~~

Heatup and Cooldown Pressure-Temperature 1 imit Curves Revision 0

b.

Figures 5-1 through 5-4 define all of the cbove limits for ensuring prevention of nonductile failure for the Byron Unit I reactor vessel. The data points for the heatup and cooldown pressure-temperature limit curves shown in Figures 51 through 5-4 are presented in Tables 5-1 through 5-4, respectively.

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 :ne initial RTNDT of the material stressed by the boltup. Therefore, since the most limiting initial RTNDT value is 60 F (closure head flange), the reactor vessel can be bolted up at this temperature.

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

1 5-3 M ATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP-5933 LIMITING ART VALUES AT 16 EFPY: 1/4T.100'F 3/4T. 92'F 2500 -

o +ii -

_ p......... l i', l l

, , i g i,i! ! ' '

._ 2250 ' '

- r' I I l' j i j

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LEAE TIST L 3 NIT { j l , ,, ,

h f I 1 .I I 1 2000 '

/ / ' '

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s .

l .

l l m 1500 / /

m l 1

  • I /

b 1250 , i ACCEPTABLE Z ~"

L

/ /

0PERATION - _.

s' A, ... i 1000 WEATUP RATB * ' ' ' ! '

UP 79 888 F/B.. j j

' \

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Q /

ri

/

e N

't FOR 100 F/Br. .'

a 750 ' ' '

m ll

,e i i I 500 e

g '

o N N 1 i i i 250 ' i I

u 7 155.Aliel':r!!!!

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iii,

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0 50 100 150 200 2$0 3d0 3$0 4d0 450 500 Moderator Temperature (Deg.F) 1 FIGURE 5-1 Byron Unit i Reactor Coolant System Heatup Limitations (Heatup Rate of 100*F/hr)

Applicable to 16 EFPY Using 1989 Appendix G Methodology (Without Margins of for Instrumentation Errors)

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

5-4 I

M ATERIAL PROPERTY BASIS i

LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP-5933 )

LIMITING ART VALUES AT 16 EFPY; l/4T.100 F 3/4T. 92*F i

2500 -

i ! i qs.........! , l;' i;i

- , - i1 - i , 1 1 me 2250 , .

1 , . ,

! +

i ii l'i i6 i l

i W3 I  ! !

o., 2000 ,/ .,i, ,-; , s l

/

j i

l i , i

{

c 1750 _____

u n i c e r y , , , 3, / l u

-g

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f i l

i 1

1

,, 1500 ,

m l

[ 1250 ,/

. , c e s , , , , i. ,

0PERA7 ION 1000 i i e

o 750 -- E!Is E""

eg F/Br.

~

e

  • '~ '

C 500 --- as I' o

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l cc 250 N aaing ,

i O

Temp l

l, l

I I I I  ! l 1

- i

' ' ' ' ' ' i '

0 0 50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F)

FIGURE 5 2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100'F/hr) Applicable to 16 EFPY Using 1989 Appendix G Methodology (Without Margins for Instrumentation Errors)

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0 m

.,- s TABLE 5-1 Byron Unit 1 Heatup Data at 16 EFPY Using 1989 App. G Methodology (Without Margins of for Instrumentation Errors) lHeatup Curves Configuration #: 1346630033 100 Heatup Critical. Limit Leak Test Limit T P T P T P 60 0 233 0 212 2000 60 531 233 531 233 2485 65 531 233 531 85 531 233 531 90 531 233 531 95 531 233 531 100 531 233 531 105 531 233 531 110 -531 233 531 115 531 233 532 120 532 233 535 125 535 233 539 130 539 233 546 135 546 233 554 140 554 233 563 145 563 233 574 150 574 233 587 l

155 587 233 602 j 160 602 233 618 165 618 233 637 -

170 621 233 657 175 621 233 679 180 621 233 703 180 679 233 729 185 703 235 758 190 729 240 789 195 758 245 822 200 789 250 859 205 822 255 898 210 859 260 940 215 898 265 985 220 940 270 1034 225 985 275 1087 230 1034 280 1143 235 1087 285 1204 240 1143 290 1269 245 1204 295 1339 250 1269 300 1414 255 1339 305 1494 260 1414 310 1580 265 1494 315 1672 Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

-v TABLE 5-1 (Continued)

Byron Unit 1 Heatup Data at 16 EFPY Using 1989 App. G Methodology (Without Margins of for instrumentation Errors)

Heatup Curves Configuration #: 1346630033  ;

100 Heatup Critical. Limit Leak Test Limit l I

T ' T P T P 270 i .#.0 320 1770 275 1672 325 1876 280 l70 330 1988 285 1876 335 2108 290 1988 340 2236 295 2108 345 2372 300 2236 305 2372 Hemtup and Cooldown Pressure-Temperature Limit Curves Revision 0

TABLE 5-2 l Byron Unit 1 Cooldown Data at 16 EFPY Using 1989 App. G Methodology (Without Margins of for instrumentation Errors) l Cooldown Curves Configuration #: 1346630033 Steady State 25F 50F 100F T P T- P T P T P 60 0 60 0 60 0 60 0 60 573 60 526 60 478 60 380 j 535 65 488 65 391 )

65 582 65 70 591 70 545 70 499 70 404 75 601 75 556 75 511 75 418 80 612 80 568 80 523 80 433 1 85 621 85 581 85 537 85 448 90 621 90 594 90 551 90 466 95 621 95 609 95 567 95 485 100 621 100 621 100 584 100 505 105 621 105 621 105 603 105 527 110 621 110 621 110 621 110 550 115 621 115 621 115 621 115 576 120 621 120 621 120 621 120 603 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 155 621 155 621 155 621 155 621 160 621 160 621 160 621 160 621 165 621 165 621 165 621 165 621 170 621 170 621 170 621 170 621 l 175 621 175 621 175 621 l 180 621 180 621 l 180 1123 180 1123 l 185 1173 190 1226 195 1284 200 1345 205 1412 4

210 1483 215 1559 220 1640 225 1728 230 1821 .

235 1921 240 2029 245 2143 250 2266 255 2397 Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

5-8

'~

M ATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL FORGING 5P-5933 LIMITING ART VALUES AT 16 EFPY: 1/4T.100 F 3/4T, 92*F 1

1 2500 .

i , i>> -

4 i - i > r i 1

- ] ... s e'ni j  ; .

,, i: l  ; l

" 2250 I ' '

- /  ! ,  ! ' .  !' ,

)

(((( LBAE TERT LIN87 { l l ,,,;

ti I i !11 i  ! t . I i

  • 2000 'l

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1000 i r \

ac i / / < calf. LINIT

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.= l :l w 500 . .

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Q 's i i I l

Tu ! l i I t

....'...,,..'..,..'.'n 1 1

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N 250 soi* . u_

a ,Ta

, ' .*, , 4. :.:.:.

. .. :1!! . .."1 . :.:,'i".!. 1 .. '. ;.;,. ,,, t:

  • 0 0 50 100 160 200 2$0 380 3$0 400 450 500 Moderalor Temperature (Deg.F)

FIGURE 5-3 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100'F/hr) I Applicable to 16 EFPY Using 1996 Appendix G Methodology (Without Margins of for i instrumentation Errors)

Hestup and Cooldown Pressure-Temperature Limit Curves Revision 0

54 L M ATERI A L PROPERTY B ASIS .

LIMITING MATERIAL: INTERMEDIATE SHELL FORGING SP 5933 LIMITING ART VALUES AT 16 EFPY: -- 1/4T.100'F -

3/4T. 92'F

< 2500.' , , .. i,- ' i

^

q ..:.,,. ; l I

l l

! l ime 7250

,i l , , .

I I  ! i W3 a 2000 ^

, l

[

l l i

i l

o '1750 vMAccrPTABLE s y (((( OPERATION I 5'1500

= '

i

'[1250 / 4ce: Prest 0FER4 TION A ,

i T

1000 l i i

'i

= c0 6....

e 750 --

~

  • )U' j

-G p- , -

c = 500 -  :  !! '

u  :: see - - 1 .

- l  !'

'l' e 250 'N, 8*uP  ! , ll,'

o Te ' i '

i , i 0 l 0 50 100 150 200 250 300 350 400 450 500 Moderalor Temperature (Deg.F)

FIGURE 5-4 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100*F/hr) Applicable to 16 EFPY Using 1996 Appendix G Methodology (Without Margas of for lastrurnentation Errors)

- Heatup and Cooldown Fassure Temperature Limit Curves Revision 0

5-10 TABLE 5-3 Byron Unit i Heatup Data at 16 EFPY Using 1996 App. G Methodology (Without Margins of for Instrumentation Errors)

Heatup Curves Configuration #: 303197761 __

100 Heatup Critical. Limit Leak Test Limit T

T P T P P J 60 0 225 0 204 2000 60 587 225 587 225 2485 65 587 225 587  !

85 587 225 587 90 587 225 587 95 587 225 587 100 587 225 587 105 587 225 587 110 587 225 587 115 587 225 588 120 588 225 591 125 $91 225 596 130 5% 225 602 135 602 225 611 140 611 225 622 145 621 225 634 150 621 225 648 155 621 225 665 160 621 225 683 165 621 225 703 170 621 225 725 175 621 225 750 180 621 225 777 I80 750 230 806 185 777 235 838 190 806 240 872 195 838 245 910 200 872 250 950 205 910 255 994 210 950 260 1041 215 994 265 1092 220 1041 270 1147 225 1092 275 1206 230 1147 280 1269 1 235 1206 -285 1338

{

240 1269 290 1411 245 1338 295 1490 250 1411 300 1575 255 1490 305 1666 260 1575 310 1764 I Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

5-11 )

l TABLE 5-3 (Continued)

Byron Unit i Heatup Data at 16 EFPY Using 1996 App. G Methodology (Without Margins of for Instrumentation Errors)

Heatup Curves Configuration #: 303197761 100 Heatup Critical. Limit Leak Test Limit T P T P T P 265 1666 315 1869 270 1764 320 1982 275 1869 325 2104 280 1982 330 2234 285 2104 335 2374 290 2234 295 2374 l

l l

/

l 1

I Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

om TABLE 5-4 Byron Unit 1 Cooldown Data at 16 EFPY Using 1996 App. G Methodology (Without Margins of for instrumentation Errors)

Cooldown Curves Configuration #: 303197761 l Steady State 25F 50F 100F T P T P T P T P 60 0 60 0 60 0 60 0 60 613 60 561 60 509 60 402 65 621 65 572 65 520 65 414 70 621 70 582 70 531 70 427 75 621 75 594 75 544 75 442 80 621 80 607 80 557 80 458 85 621 85 620 85 572 85 475 90 621 90 621 90 588 90 494 95 621 95 621 95 605 95 514 100 621 100 621 100 621 100 535 105 621 105 621 105 621 105 559 110 621 110 621 110 621 110 584 115 621 115 621 115 621 115 611 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 155 621 155 621 155 621 155 621 160 621 160 621 160 621 160 621 165 621 165 621 165 621 165 621 170 621 170 621 170 621 170 621 175 621 175 621 175 621 180 621 180 621 180 1207 180 1205 185 1261 190 1319 195 1382 200 1449 205 1521 210 1599 215 1683 220 1773 225 1869 230 1973 235 2085 240 2205 245 2334 250 2473 4

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

~ -

61 6l REFERENCES i

1 ' Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials". U.S.

- Nuclear Regulatory Commission, May,1988.

2 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", Federal Register. Volume 60.

No. 243, dated December 19,1995.

~31 1989 ASME Boiler and Pressure Vessel (B&PV) Code,Section XI. Appendix G. " Fracture Toughness Criteria for Protection Against Failure".

4 CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI .  !

Consulting, March 1996.

5 WCAP-14824, Revision 2, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld MetalIntegration For Byron and Braidwood", T. J. Laubham, et al., November 1997.

6 1989 Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Matenal for Vessels".

7 WCAP-15123, " Analysis of Capsule W from the Commonwealth Edison Co. Byron Unit 1 Reactor Vessel Radiatioa Surveillance Program" T. J. Laubham, et al., November 1998.

8. WCAP-14040-NP-A, Revision 2, " Methodology used to Develop Col' d Overpressure Mitigating  ;

System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D. Andrachek, et al., January 1996.

9. Letters CAE-97-231, CCE-97-314," Comed Response to NRC Question to WCAP-14940, WCAP-14970 and 14824 Rev. 2", From C.S. Hauser to Mr. Guy DeBoo (of Comed), Dated January 6,1998.
10. Babcock & Wilcox drawing numbers 184557E, Rev. 2; I85266E, Rev. 2; I 85297E, Rev. 2; 185328E, Rev. 2; " Reactor Vessel Longitudinal Section".

I 1. ASME Boiler and Pressure Vessel Code,Section XI, " Rule for Inservice Inspection of Nuclear Power Plant Components", Appendix G," Fracture Toughness Criteria for Protection Against Failure", December 1995.

12. I.S Raju and J.C. Newman, Jr., " Stress Intensity Factor Influence Coefficients for Internal and External Surface Cracks in Cylindrical Vessels", in Aspect of Fracture Mechanics in Pressure Vessels and Piping, ed. S.S. Palusamy and S.G. Sampath, PVP-Volume 58, ASME 1982.

References Revision 0

6-2 f i

l

13. NRC SER Dated January 21,1998. " Byron Station. Units 1 and 2 and Braidwood Station. Units 1 and 2, Acceptance For Referenemg of Pressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, and M98802), From R.M. Krich to 0. D, Kingsley.
14. Comed Letter to U.S. Regulatory Commission, " Response to Additional Information Regarding Reactor pressure Vessel", From R.M. Knch, Dated September 3.1998.
15. NDIT No. MSD-98-044, "Best Estimate Chemistry Values for Reactor Pressure Vessel Beltime Weld Heat Number 442002", Dated December 1998.

I hh Revision 0 1

Attachment J

)

Byron Station j

)

WCAP-15178, Rev. O," Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation" l

1 l

i 1

e p:W9byttrsWOO95. doc

i. t
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