ML20207H807

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Rev 0 to WCAP-15178, Byron Unit 2 Heatup & Cooldowm Limit Curves for Normal Operations
ML20207H807
Person / Time
Site: Byron  Constellation icon.png
Issue date: 06/30/1999
From: Christopher Boyd, Laubham T, Trombola D
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20207H753 List:
References
WCAP-15178, WCAP-15178-R, WCAP-15178-R00, NUDOCS 9907210217
Download: ML20207H807 (54)
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15178 Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubham June 1999 Work Performed Under Shop Order C8QP-139A Prepared by the Westinghouse Electric Company LLC for the Commonwealth Edison Company Approved:

C. H. Boyd, ManagW Equipment & Materials Technology Approved: -

M D. M. Trombola, hfanager Mechanical Systems Integration Westinghouse Electric Company LLC Energy Systems P.O. Box 355 Pittsburgh, PA 152.M)4355 C1999 Westinghouse Electric Company LLC All Rights Reserved c

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TABLE OF CONTENTS i 1

LIST OF TABLES.. .... .. . . .. . . . . . . . . .iv

. LIST OF FIOURES.. ..... . . . . . . . . . . . . . . . . . . . . . . .. . . . vi l

i PREFACE ......................................................... . . . vii

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EXECUTIVE S UMMARY . .. ..... ... .. ........ ...... . .. ......... .... . ... . ... .. . . . . . . . . . . . . . . . viii l

1 INTRODUCTION . ..... . .. . .......... . . .... .. .. . . . . . . . . . . . . . . . . . . .1-1 2- PURPOS E . . . .. . .. . . ... . . ... . . ... .. . . . . . . . . . . . . . . . . . . . . . . . ...... . . 2-1 1

3 CRITERIA FOR ALLOWABLE PRESSURE TEMPERATURE RELATIONSHIPS.. . . . . 3-1 4

CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ..... . . .. . . . .. .. . 4 1 5' HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES... . . .. . 5-1 i i

I 6 REFEREN CES . . . . ... . .. ... . . . . . . . .. . . . . .. . ... . .... . .. . .. . .. . . . .. . . . . . ..... . . . . . . .. . 6-1 i

l Revision 0

r iv l LIST OFTABLES l

l Table 4-1 Summary of the Peak Pressure Vessel Neutron Fluence Values at 16 EFPY used 2

! for the Calculation ofART Values (n/cm , E > 1.0 MeV) . . .44

! Table 4-2 Calculated Integrated Neutron Exposure of the Byron Unit 2 Surveillance Capsules Tested to Date. . . . . ...... . . . .4-4 Table 4-3 Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program ..... . . .. . . . . . . . . .. . .4-5 Table 4-4 Calculation of Best Estunate Cu and Ni Weight Percent for the ByTon Unit 2 Forging Matenals....... ..... .. ... . . . . .. ... . . . . . . . . . . ... .4-6 Table 4-5 Calculation of the Average Cu and Ni Weight Percent for the Byron Unit 2 Surveillance Weld Material Only (Heat # 442002). . . . . . . .. . . .4-6 Table 4-6 Calculation of Best Estimate Cu and Ni Weight Percent Values for the Byron Units 1 & 2 Weld Material (Using Byron 1 & 2 Chemistry Test Results). . . ... ... . 4-7 Table 4-7 Reactor Vessel Beltline Material Umrradiated Toughness Properties.. . . . ... .. .4-8

'lable 4-8 Calculation of Chemistry Factors for Byron Unit 2 using Surveillance Capsule Data .. . . .. .. ... .. . . . . . . . . . .... ... . .. . . . . . . . . . . . . . . . . . . 4-9 Table 4-9 Summary of the Byron Unit 2 Reactor Vessel Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1. . 4-11 Table 4-10 Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 16 EFPY Heatup / Cooldown Curves... .. . ... .. .. . .4-12 Table 4-11 Calculation of the ART Values for the 1/4T Location @ 16 EFPY.. . . . . ...4-13 Table 4-12 Calculation of the ART Values for the 3/4T Location @ 16 EFPY.. . . . . . . . . ...4-14 Table 4-13 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location... . . 4-15 Table 5-1 Byron Unit 2 Heatup Data at 16 EFPY Using 1989 App. G Methodology (Without Margins for lastrumentation Errors). .. ....... . . .. . . . . . . ... .. . .5-5 Table 5-2 Byron Unit 2 Cooldown Data at 16 EFPY Using 1989 App. G Methodology .

(Without Margins for Instrumentation Errors) .... .. . ... ........ . . .. . .. . . . . . . . 5 -7 Revision 0

v LIST OFTABLES Table 5-3 Byron Unit 2 Heatup Data at 16 EFPY Using 1996 App. G Methodology (Without Margins for Instrumentation Errors) .. . . . .. . .5-10 Table 5-4 Byron Unit 2 Cooldown Data at 16 EFPY Using 1996 App. G Methodology (Without Margins for Instrumentation Errors) . . . .5-11 Table 5-5 Byron Unit 2 Heatup Data at 16 EFPY Using Code Case N-588 vs.1996 App. G Methodology (Without Margins of for Instrumentation Errors) . . . _ . . .5-14 Table 5-5 Byron Unit 2 Cooldown Data at 16 EFPY Using Code Case N-588 vs.1996 App. G Methodology (Without Margins of for Instrumentation Errors) . . . .5-15 Revision 0

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l.- LIST OF FIGURES l Figure 5-1 Byron Unit 2 Reactor Coolant System Heatup Limitations Ricatup Rate of 100*F/hr)

Applicable to 16 EFPY Using 1989 App. G Methodology (Without Margins for Instrumentation Errors)... . . . 5-3 l Figure 5-2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates l

of 0,25,50 and 100'F/hr) Applicable to 16 EFPY Using 1989 App. G Methodology (Without Margins for 1:wtrumentation Errors)... . .5-4 i

Figure 5-3 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100 F/hr)

Applicable to 16 EFPY Using 1996 App. G Methodology (Without Margms for Instrumentation Errors)...... . .. . . . . . . .. .. ..5-8 Figure 5-4 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100'F/hr) Applicable to 16 EFPY Using 1096 App. G j Methodology (Without Margms for Instrumentation Errors).. .. . . . . . . . 5-9 l

Figure 5-5 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100'F/hr) Applicable to 16 EFPY Using Code Case N-588 vs.1996 Appendix G with Axial ART (Without Margms of for Instrumentation Errors).. . . ...5-12 Figure 5-6 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and100'F/hr) Applicable to 16 EFPY Using Code Case N 588 vs.1996 Appendix G with Axial ART (Without Margms of for Instrumentation Errors) . ... .5-13 l

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PREFACE his report has been technically reviewed and verified by:

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- Reviewer: Ed Terek 4

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EXECUTIVE

SUMMARY

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l The purpose of this report is to generate pressure-temperature limit curves for Byron Unit 2 for normal l 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 l Guide 1.99, Revision 2 is used for the calculation ofAdjusted Reference Temperature (ART). This report j also presents pressure-temperature limit curves that follow the requirements of the 1996 Addenda to Appendix G, including the circumferential flaw methodology, for calculating the stress intensity factors.

The 1/4T and 3/4T values are summarized in Table 4-13 and were calculated using the circumferential weld WF-447 (Heat 442002) and nozzle shell forging 4P-6107 (The limiting material when axial flaws are not postulated in the circumferential welds, Code Case N-588). The pressure-temperature limit cun'es 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 5-6.

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1-1 1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTm (reference nil-ductility

' temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTm of the limiting mrial in the core region of the reactor vessel is determined by using the unirt=AatM reactor etal material fracture toughness properties, estimating the radiation-induced ARTm.

and addmg a margin. The unirradiated RTm is designated as the higher of either the drop weight nil-ductility transition temperature (NDTf) 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.

RTm increases as the material is exposed to fast-neutron radiation. Therefore, to fmd the most limiting RTm t any a time period in the reactor's life, ARTer due to the radiation exposure associated with that time period must be added to the unirr=AntM RTa(IRTwr). The extent of the shift in RTm is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Rag"I-*~y Commission (NRC) has published a method for predicting radiation embnttlement in Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials"IU Regulatory Guide 1.99, Revision 2, is used for the calculation ofAdjusted Reference Temperature (ART) values (IRTer + ARTer + marsms for uncertamties) 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 r=Amtw= survmil== program, Babcock and Wilcox Co. supplied Westmghouse with sections of SA508 Class 3 forging material used in the core region of the Byron Station Unit No. 2 reactor pressure vessel (Speci6cally from forging [49D330/49C298]-1-1). Also supplied was a weldment made with non-copper coated weld wire using the automatic sub-arc welding process (Weld wire heat # 442002 Linde 80 flux, lot number 8064, which is identical to that used in the actual fabrication of the intermedate to lower shell girth weld of the pressure vessel).

Introduction Revision 0

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2 PURPOSE I

The Commonwealth Edison Company contracted Westinghouse to analyze surveillance capsule X from the  !

Byron Unit 2 reactor vessel. As a pan of this analysis Westinghouse generated new heatup and cooldown i curves for 16 EFPY using the methodology from WCAP-14040-NP-A and the 1989 and 1996 ASME B&P Vessel Code,Section XI, Appendix G and from Code Case N-588. 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 Pan 50, Appendix Gr21 The purpose of this report is to present the calculations and the development of the Commonwealth Edison Company Byron Unit 2 heatup and cooldown curves for 16 EFPY This report documents the calculated

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i adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2 t u, 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, j Byron Unit 2, Braidwood Unit I and Braidwood Unit 2 surveillance programs has been irtegrated. Note that Byron Unit 2 surveillance weld is identical to the surveillance weld (Heat No. 442002) at Byron

' Unit 1, and the Braidwood Units'survedlance weld is identical to the nozzle shell forging to intermediate shell fonpng ginh weld (Heat No. 442011) at Byron Unit 2. Per WCAP-15180071, all the surveillance data has been determmed to be credible, i

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Purpose Revision 0

3-1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS i

Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements"Pl 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 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 j Code forms the basis for these requirements.Section XI, Division 1, " Rules for Insenice Inspection of Nuclear Power Plant Components", Appendix Gm, contains 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, Ki, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Ki., for the metal temperature at that time. Ki. is obtamed from the reference fracture toughness curve, defmed in Appendix G of the ASME Code,Section XI. The Kr. curve is given by the following equation:

L. = 26.78 + 1.233

  • ctoo m r-ar u m g3)

I where, Ki. = reference stress intensity factor as a function of the metal temperature T and the metal  !

reference nil-ductility temperature RTum Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C*Kw+L,<L. (2) where, Kw= stress intensity factor caused by membrane (pressure) stress j l

Ki, = stress intensity factor caused by the thermal gradients j' Ks. = function of temperature relative to the RTum of the material i C= 2.0 for Level A and Level B service limits l l C= 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical l

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l Cntena For Allowable Pnssure-Temperature Relationships Revision 0 L

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3-2 At any time during the heatup or cooldown transient, Kr. is determmed by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTwr, 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, Kit, 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 i 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 because control of the cooldown e- ~ is based on the measurement of reactor coolant temperature, whereas the limiting pressure is ndent on the matenal temperature at the tip of the assumed flaw. During cooldown, the 1/4T

.a. socation is at a higher a ..,~.wre 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 L,...,~.ere, the AT (L. are) 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 Kii, 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 eliminates 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 fmite 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 assummg the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that allesiate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the Ki , for the 1/4T crack during heatup is lower than the Ki , 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 nresses and lower Ki. values do not offset each other, and the pressurce..,miare 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.

LJe, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and fmite heatup rates is obtamed The second portion of the heatup analysis concems the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during I heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses Criteria For Allowable Pressure-Temperature Relationships Revision 0

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present. These thermal stresses are daaa-daat on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with ]

' increasing heatup rates, each heatup rate must be analyzed on an indisidual basis.

Following the generation of pressure-temperature curves for both the steady state and fmite heatup rate situqions, the fmal limit curves are produced by constructing a composite curve based on a point-by-point i

comparison of the steady-state'and finite heatup rate data. At any given temperature. the allowable

_ pressure is taken to be the lesser of the three values taken 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 i

regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirrad=wl RTm 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 ByTon Unit 2 reactor vessel.

De limiting unirradiated RTwar o f 30'F (Table 4-7) occurs in the vessel flange of the Byron Unit 2 reactor vessel, so the muumum allowable L .psure of this region is 150'F at pressures greater than 621 psig.

I 1996 Addaad= to ASME Saetian XI. Annaadiv G Mathadalonv "3 In 1996 Appendix G was revised to incorporate the most recent clastic solutions for Ki due to pressure and radial thermal gradients. The new solutions are based on finite element analyses for inside surface flaws 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 :t 21,'

His revision now provides consistent computational mad =le 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 Kidue 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 relatively small in terms of absolute allowable pressure, the benefits are substantial because even a small increase in the allowable pressure can be a significant percentage increase in the operatmg window at .

l relatively low temperatures. Implementmg this revision results in an economic and potential safety benefit l (less likehhaad oflifting LTOP relieving devices) with no reduction in vessel integrity; i.e. as an input to LTOP set points, the improvement in steady state maxunum allowable pressure for B3 Ton Unit 2 at 60*F is  ;

. -40 psig.  ;

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Criteria For Allowable Pressure-Temperature Relationships Revision 0

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

G-2214.1 Membrane Tension:

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

M. 1,85 for d < 2,

=

M. =.- 0.926E for 2s d s 3.464, M.' = '3.21 for E > 3.464 Sinularly, W for an outside surface flaw is given by:

A = l'.77 for d < 2, 1

M. = 0.8938 for 2s d s 3.464, M.: =. 3.09 for d > 3.464 ,

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

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G-2214.3 Ra&alThermal Gradient:

The maxunum Ki produced by ra&al thermal gradient for the postulated inside surface defect of G-2120 is Kii = 0.953x10 ' x CR x t", where CR is the cooldown rate in 'F/hr., or for a postulated outside surface defect, Ki, = 0.753x10 x HU x t", where HU is the heatup rate in 'F/hr.

The through-wall temperature difference associated with the maximum thermal Ki can be deternuned from Fig. G-2214-1. The ^. . pron.it at any radial & stance from the vessel surface can be determined from Fig.

G-2214-2 for the maximum thermal Ki ,

(a); The maximum thermal Ki relationship and the ^_.g.Lare relationship in Fig. G-2214-1 are applicable only for the con &tions given in G-2214.3(a)(1) and (2).

(b) Alternatively, the Ki for ra&al thermal gra&ent can be calculated for any thermal stress distribution and at any specified time during cooldown for a %-thickness inside surface defect using the relatwinabp:

Critena For Allowable Pressure Temperature Relationships Revision 0

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L, = (1.0359Co + 0.6322Ci + 0.4753C2 + 0.3855C3)

  • 5 (4) 1 or similarly, Krr during heatup for a %-thickness outside surface defect using the relationship:

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M, = (1.043Co + 0.630Ci + 0.481C2 + 0.40IC3)

  • M (5) t 'where the coefficients Co, Ci, C2 and C 3are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

a(x) = Co + Ci(r / a) + C2(r / a)2 + C3(r / 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 maxunum crack depth.

Note, that equations 3,4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-L..,, &are (P-T) limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T cun e methodology is unchanged from that described in WCAP-14040N Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above. Per Reference 13, the NRC has reviewed and accepted this methodology for Byron Unit 2 in January of 1998.

Code r=- N-588: Circumferwi=1 Welds:

In 1997, ASME Section XI, Appendix G was revised to add methodology for the use of circumferential flaws when considermg circumferential welds in developing pressure-temperature limit cun'es. This change was also implemented in a separate Code Case, N-588.

"Ihe earlier ASME Section XI, Appendix G approach mandated the postulation of an axial flaw in circumferential welds for the purposes of calculatmg pressure-temperature limits. Postulating the Appendix G reference flaw in a circumferential weld is physically unrealistic because the length of the

.4...cc flaw is 1.5 times the vessel thickness and is much longer than the width of the vessel girth welds.

In addition, historical expenence, with repair weld indications found during pre-service inspection and data taken from destructive exammation of actual vessel welds, confirms that any flaws are small, lammar in nature and are not oriented transverse to the weld bead orientation. Because of this, any defects potentially ,

introduced during fabrication process (and not detected during subsequent non-destructive exammations)

Critena For Allowable Pressure-Temperature Relationships Revision 0 I l

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-r should only be oriented along the direction of the weld fabrication. Thus, for circumferential welds, any postulated defect should be in the circumferential orientation.

= The revision to Appendix G now climmates additional conservatism in the assumed flaw orientation for circumferential welds. 'The following revisions were made to ASME Section XI, Appendix G:

G-2214.1 Membrane Tension...

The Ki corresponding to membrane tension for the postulated circumferential defect of-2120 is Kw = M. x (pR,l t) where, M. for an inside surface flaw is given by:

M. .= 0.89 for E < 2, M. = '0.443 d for 2s 8 s 3.464,

%= _1.53 for E > 3.464 Simdarly, % for an outside surface flaw is given by:

M. = 0.89 for E < 2, M. = 0.443E for 2s 8 5 3.464, M. = 1.53 for 8 > 3.464 Note, that the only change relative to the OPERLIM computer code was the addition of the constants for M. in a cire. weld limited condition. No other changes were made to the OPERLIM computer code with reprd to P-T calculation methodology. As stated previously, the P-T curve methodology is unchanged from that described in WCAP-14040 1'l Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

Criaena For Allowable Pressure-Temperature Relationships Revision 0

4-1 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: 1 ART - InitialRTar + ARTwr + Margin (7)

Initial RTurr is the reference temperature for the unitradiated material as defmed in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Codef 'l If measured values ofinitial RTmyr for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ARTmn is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows:

A RTar = CF

  • f****** gg; To calculate ARTmyr 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,,%* e* (9) 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 resultant fluence is then placed in Equation 8 to calculate the ARTmyrat the specific depth.

The Westinghouse Radiation Engmeering and Analysis group evaluated the vessel fluence projections and the results are presented in Section 6 ofWCAP-15176"1 The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with the methods presented in WCAP-14040-NP A,

" Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"I'l. Table 4-1, herein, contains the calculated vessel surface fluence values along with the Regulatory Guide 1.99, Revision 2,1/4T and 3/4T calculated flucaces used to calculate the ART values for all beltline materials in the Byron Unit 2 reactor vessel. Additionally, the calculated surveillance capsule fluence values are presented in Table 4-2.

Ratio Procedure and To...midere 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 ARTmn of the weld materials for differences in copper / nickel content. This adjustment is performed by multiplying the ARTmyr by the ratio of the vessel chemistry factor to the surveillance material chamictry factor. The adjusted ARTurr values are then used to calculate the chemistry factor for the vessel materials.

Calculation of Adjusted Reference Temperature Revision 0

42_

From NRC Industry Meetags on November 12,1997 and February 12,13 of 1998, procedural guidelines were presented to adjust the ARTwor for temperature differences when using surveillance data from one vessel applied to another vessel. The followmg guidance was presented at these industry meetings:

Irradiation temperature and fluence (or fluence factor) are first order emironmental variables in -l assessing irradition damage... To account for differences in temperature between sun eillance 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 l'Fincreasein ARTwor.

For capsules with irradiation temperature of T% and a plant with an irradiation temperature of T,i , an adjustment to normalize ARTwor. ,4 to T,i is made as follows:

n .=.

Temp. Adjusted ARTwor = ARTwor.,,,,,,,,4 + 1.0*( Tw. - T,i ) (10)

De irradiation L..,, ,4 ares from Byron Units 1 & 2 and Braidwood Units 1 & 2 are presented in WCAP-14824, Revision 2. The average irradiation temperature from each of the four Units and operating cycles in question is 553*F. Herefore, no .#4.re adjustment is required.

Chemistry Factor:

The chemistry factor is obtmaad from the tables in Regulatory Guide 1.99, Revision 2 using the best estamate average copper and nickel content as reported in Tables 4 4 through 4-7. He chemistry factors were also calculated using Position 2.1 from the Ret;ulatory Guide 1.99, Resision 2 using all available surveillance data. For weld WF-447, the surveillam e data credibility assessment in Reference 17, d:t deed the Byron Unit 2 weld metal data were credible with and without the surveillance data from Byron Unit 1. Ais assessment also calculated the vess 3, weld (including temperature and chemistry adjustment) dwidstry factor using Byron Unit 2 stand alone data and it was determined to be 64.6*F.

Position 2.1 chemistry factors are calculated in Table 4-8.

)

Calculacion of Adjusted Reference Temperature - Revision 0

b 4-3 Explanation of Marnin Term:

When there are %c or more credible surveillance data sets'M available for Byron Unit 2, Regulatory j Guide 1.99 Rev. 2 (RGl.99R2) Position 2.1 states "To calculate the Margin in this case, use Equation 4; I the values given there for e6 may be cut in half". Equation 4 from RGl.99R2 is as follows:

M = 2da' + o?

1

) 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 2. then o nis equal

~

to 0*F. On the other hand, if the initial RTm values were not measured, then a generic value of 17'F I (base metal and weld metal) would have been required to be used for a .

l Standard Deviation for ARTm Margm Term, a6 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 c6 need not exceed 0.50 times the mean value of ARTwor." The mean value of ARTm is defined in RGl.99R2 by Equation 2 and defined herein by Equation 8.

Per RGl'99R2 Position 2.1, when there is credible surveillance data, c6 si taken to be the lesser of %

ARTwor or 14*F (28'F/2) for welds, or 8.5'F (17'F/2) for base metal. Where ARTm gain a is defined t

herem by Equation 8.

Summary of the Margm Term Since oi is taken to be zero when a heat-specific measured value ofinitial RTwor are available (as they are in this case), the total margm term, based on Equation 4 of RGl.99R2, will be as follows:

. Position 1.1: Lesser of ARTwor or 56'F for Welds Lesser of ARTwor or 34*F for Base Metal i

e Position 2.1: Lesser of ARTwor or 28'F for Welds Lesser of ARTwor or 17'F for Base Metal l

Calculation of Adjusted Reference Temperatus, Revision 0

I i l 4-4 j I

TABLE 4-1 i l

Summary of the Peak Pressure Vessel Neutron Fluence Values 2

at 16 EFPY used for the Calculation ofART Values (n/cm , E > 1.0 MeV)

Azimuth Surface %T %T Nozzle Shell Forging 4P-6107 2.54 x 10" 1.53 x 10" 5.50 x 10"

'f Intermediate Shell Forging 1.00 x 10" 6.00 x 10" 2.17 x 10"

[49D329/49C297]-1-1 Lower Shell Forging [49D330/49C298]-1-1 1.00 x 10" 6.00 x 10" 2.17 x 10" Intermediate to Lower Shell Forging Cire. 9.86 x 10" 5.92 x 10" 2.14 x 10" Weld Seam WF-447 (Heat 442002)

> Nozzle Shell to Intermediate Shell Forging 2.54 x 10" 1.53 x 10" 5.50 x 10" l Cire. Weld Seam WF-562 (Heat 442011) 2 Note: All remauung vessel materials are below I x 10" n/cm , E > 1.0 MeV TABLE 4-2 Calculated Integrated Neutron Exposure of the Byron Unit 2 Surveillance Capsules Tested to Date Capsule Fluence U 4.05 x 10" n/cm2 , (E > 1.0 MeV)

W l.27 x 10" n/cm 2, (E > 1.0 meV)

X 2.30 x 10" n/cm2 , (E > 1.0 MeV)

Calculation of Adjusted Reference Temperature Revision 0

r 4-5 Contained in Table 4-3 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline materials. These measured shift values were obtamed using CVGRAPH, Version 4.lM, which is a symmetric hyperbolic tangent curve-fitting program.

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 U -3.8 F

[49D330/49C298]-1-1 W 3.65'F (Tangential Orientation) X 15.75'F Intermediate Shell Forging U 19.76'F

[49D330/49C298]-1-1 W 31.88'F (Axsal Orientation) X 38.91*F Surveillance Program U 8.44*F Weld Metal 'W 28.88'F (Heat # 442002) X 54.0l'F Heat Affected Zone U 6.74'F W 30.44*F X 34.22 F Notes..

(a) From capsule X analysis resultst g Calculation of Adjusted Reference Temperature Revision 0

.I

4-6 Table 4-4 contains the calculation of the best estimate weight percent copper and nickel for the Byron Unit 2 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 .2 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 RTuorof the beltline materials and vessel flanges. The weight percent values of Cu an ' di given in Table 4-7 were used to generate the calculated chemistry factor (CF) values based on Tables 1 ad 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 4-4 M Calculation of the Best Estimate Cu and Ni Weight Percent for the Byron Unit 2 Forging Materials Intermediate Shell Forging Lower Shell Forging 49D329/49C297-1-1 49D330/49C298-1-1 Reference Cu % Ni % Cu % Ni %

Ref.16 O.01 0.70 0.05 O.71"'

Charpy YL-48*' --. --- 0.075 0.78 Charpy YT-52*) --- --- 0.073 0.78 Best Estimate Average 0.01 0.70 0.06 0.73 Note:

(a) This is the average of 4 data poitets (b) Charpy Specimens From Capsule X of Byron Unit 2 (Ref. 7).

(c) The best estimate 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 2 Surveillance Weld Material Only (Heat # 442002)

Reference Weight % Copper Weight % Nickel Ref. S O.023 0.712 Charpy YW-51*' O.031 0.70 Charpy YW-52*) 0.033 0.70 Surveillance Weld Average 0.02 O.71"'

Note:

(a) Thisis the average of 31 data points.

(b) Charpy Specimens From Capsule X of Byron Unit 2 (Ref. 7).

(c) The best estimate average was rounded per ASTM E29, using the " Rounding Method" 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)")

Chemistry Type Reference Weight % Copper Weight % Nickel Ref. 5 0 024 0.70 B&W WQ: BAW-2261 Ref. 5 0.031 0 46 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: BAW 2261 Ref. 5 0.029 0.65 B&W WQ (from NDIT No.

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." Ref.18 0.022 0.69 Byron 2 Surveillance Ave." Table 4 5 0.02 0.71 BEST ESTIMATE AVERAGE

  • 0.04
  • 0.63"'

E91LE (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 o that used in the reactor vessel core region girth seam (WF-i 447). The weld wire is type Linde MnMoNi (Lov Cu-P), heat number 442002, with a Linde 80 type flux, lot number 8064.

(c) The best estimate chemistry values were obtained usue, the" average of averages approach. In addition 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). per Ref.15.

Calculation of Adjusted Reference Temperature Revision 0

4-8 I

TABLE 4-7 Reactor Vessel Beltline Material Unitradiated Toughness Properties Material Description Cu(%) Ni(%) Initial RTer")

Closure Head Flange 5P7382 / 3P6407 --- 0.71 0 Vessel Flange 124L556VA1 --- 0.70 30 Nozzle Shell Forging 4P4107*) 0.05 0.74 10 Intermediate Shell Forging [49D329/49C297]-1-1 0.01 0.70 -20 Lower Shell Forging [49D330/49C298]-1-1 0.06 0.73 -20 intmnediate to Lower Shell Forgmg Cire. Weld 0.04 0.63 10 Seam WF-447 (Heat # 442002)

Nozzle Shell to Intermediate Shell Forging Cire 0.03 0.67 40 Weld Seam WF-562 (Heat # 442011)*)

Byron Unit 1 Surveillance Program 0.02 0.69 ---

Weld Metal (Heat # 442002)

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)

Notes.

(a) The initial RTervalues for the plates and welds are based on measured data per reference 5 and 9.

(b) Ber,t Estimate Cu% / Ni% and initial RTer Per Reference 5 and/or 9,15 and 18.

)

)

l i

Calculation of Adjusted Reference Temperature Revision 0 l

i

p ,

4-9 'j l

I l TABLE 4-8 i f Calculation of Chemistry Factors for Byron Unit 2 using Surveillance Capsule Data l Material Capsule Capsule f'" FF" ARTer" 4 FF*ARTer FF Lower Shell Forging U 0.405 0.749 0.0'" 0 0.561

[49D330/49C298] 1-1 W l.27 1.067 3.65 3.89 1.138 (Tangential) X 2.30 1.225 15.75 19.29 1.500 l

Lower Shell U 0.405 0.749 19.76 14.80 0.561 Forging [49D330/ W l.27 1.067 31.88 34.02 1.138 X 2.30 1.225 38.91 47.66 1.500 49C298]-1 1 SUM: 119.66 6.398 CFr.,,,, = I(FF

  • RTer) + Z( FF ) = (l19.66) + (6.398) = 18.7'F Byron Unit 1 Sury. Weld U 0.404 0.749 11.22 8.40 0.561 Material (5.61)

(Heat # 442002) X 1.57 1.125 80.22 90.25 1.266 (40. l l)(d' W 2.43 1.239 102.68 127.22 1.535 (51.34)(d)

Byron Unit 2 Sury Weld U 0.405 0.749 16.88 12.64 0.561 Material (8.44)'d)

(Heat # 442002) W l.27 1.067 57.76 61.63 1.I38 (28.88)(d' X 2.30 1.225 108.02 132.32 1.500 (54.01)(d' SUM: 432.46 6.561 CFs., w.a 442002 = I(FF

  • RTer) + Z( FF 2) = (432.46) + (6.561) = 65.9'F Notes.

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

(b) FF = fluence factor = f<a23.ci% o ,

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

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

No temperature adjustment are required.

l (c) Actual value of ARTer is -3.8. This physically should not occur, therefore for conservatism (i.e. higher chemistry factor) a value of zero will be used.

l Calculation of Adjusted Reference Temperature Revision 0

4-10 TABLE 4 Continued Calculation of Chemistry Factors for Byron Unit 2 using Surveillance Capsule Data Material Capsule ' Capsule f*) FF8 ARTer"' FFaARTrmr FF Weld Heat 442011, WF 562 U 0.3814 0.733 10 7.3 0.537 Using Braidwood 1 Surv. X 1.144 1.038 25 26.0 1.077 Data Weld Heat 442011, WF-562 U 0.3933 0.741 0 0 0.549 Using Braidwood 2 X 1.126 1.033 20 20.7 1.067 Surv. Data SUM: 54 0 3.23 CFsur, w.id 442o1: = Z(FF

  • RTmn) + Z( FF ) = (54.0) + (3.23) = 16.7*F"'

Notes.

(a)

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

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

(b) FF = fluence factor = fo.2s.o iw o I

(c) v ARTmn alues are the measured 30 fi-lb shift values taken from Appendix B of Ref. 5.  !

(d) (

The Braidwood I & 2 surveillance weld metal ART >erv alues do not require a ratio factor or temperature adjustment.

j (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. '

I i

i l

l 1

I i

Calculation of Adjusted Reference Temperature Revision 0 1

t 4 11 TABLE 4-9 Summary of the Byron Unit 2 Reactor Vessel Beltline Material 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 Intermediate Shell Forging 20.0*F ---

[49D329/49C297]-1-1 Lower Shell Forging. 37.0*F 18.7 F

[49D330/49C298]-1-1 Nozzle Shell Forging 4P-6107 ,

31.0*F ---

Intermediate Shell to Lower Shell Forging 54.0*F 65.9'F Cire. Weld Seam WF-447 (Heat 442002)

Nozzle Shell to Intermediate Shell Forging 41.0 F 16.7*F Cire. Weld Scam WF-562 (Heat 442011)

Byron Unit 1 Surveillance Program 27.0*F ---

l Weld Metal Byron Unit 2 Surveillance Program 27.0*F ---

Weld Metal i

Braidwood Unit 1 & 2 Surveillance 41.0*F ---

Program Weld Metal I

i I

1 Calculation of Adjusted Reference Temperature Revision 0

.4 4-12 Contained in Table 4-10 is the summary of the fluence factors (FF) used in the calculation of adjusted Teference temperatures for the Byron Unit 2 reactor vessel beltline materials for 16 EFPY.

TABLE 4-10 Calculation of the 1/4T and 3/4 T Fluence Factor % lues used for the Generation of the 16 EPFY Heatup/Cooldown Curves Azimuth - 1/4 T F 1/4T FF 3/4T F 3/4 T FF 8

(m/cm , E > 1.0 MeV) (n/cm'. E >1.0 MeV)

Intermediate Shell Forging 6.00 x 10" . 0.857 2.17 x 10" 0.589

[49D329/49C297] 1-1 Lower Shell Forging 6.00 x 10" 0.857 2.17 x 10" 0.589

[49D330/49C298]-1 Nozzle ShellForging 4P 6107 1.53 x 10" 0.507 5.50 x 10" 0.308 Intermediate to Dmer Shell Forging 5.92 x 10" 0.853 2.14 x 10" 0.586 Cire. Weld Seam WF-447 (Heat 442002)

Nozzle Shell to later. ShellForging 1,53 x 10" 0.507 5.50 x 10" 0.308 Circ. Weld Scam WF 562 (Heat 442011)

I 1

l 1

C-=iW 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 girth weld WF-447 is the limiting beltline materials for all heatup and cooldown cunes to be generated. The ART value associated with this material will be used in all three sets of curves. However, when generating curves for Code Case N-588 (ie. Cire. Flaw), the ART associated with the limiting axial material must also be considered to determine if this case would be more conservative or overlap the cire.

flaw curves. Contained in Table 4-13 is a summary of the limiting ARTS to be used in the generation of the Byron Unit 2 reactor vessel heatup and cooldown curves.

-TABLE 4-13 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location for 16 EFPY Material 16 EFPY 1/4 TART 3/4 TART Intermediate Shell Forging 14 4

[49D329/49C297]-1-1 Lower Shell Forging 43 24

[49D330/49C298]-1-1

- Using Surveillance Data") 12 2 CircumferentialWeld WF 447 102 73

- Using Surveillance Data 94W 77*

CircumferentialWeld WF-562 82 65

- Using Surveillance Data from 57 50 Braidwood I and 2 Nozzle Shell Forging 4P-6107 41 " 29

  • NOTES:

(a) These ART values were used to generate the Byron Unit 2 heatup and cooldown curves in 1 Figures 5-1 through 5-4. See note (b). l (b) These ART values, using the '96 App. G Methodology produced a more consen ative cun e, with no overlap, than the curves with the cire. flaw ART values using Code Case N-588 Methodology (See Figures 5-5 and 5-6),

i 1

Calculation of Adjusted Reference Temperature Revision 0 o l

51 5 L HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES i 1

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 M~ls discussed in Section 3 and 4 of this report. This approved methodology is also presented in-WCAP-14040-NP-A"l, dated January 1996.

- Figures 5-1 through 5-6 present the 16 EFPY heatup 'and cooldown cun'es (without margins for possible mstrumentation errors) for a heatup rate of 100*F/hr and cooldown rates of 0,25. 50 and 100'F/hr using i the 1989 Appendix G methodology, the 1996 Appendix G methodology and Code Case N 588 respectively.

The heatup and cooldown curves for Code Case N-588 actually are curves generated using the 1996 App.

' G methodology with the lower arialflaw ARTvalue. The reason is, these cun'es are more consen ative than the curves generated using Code Case N-588 methodology with the higher circ. flaw ART value. This is true dem.p.est the entire temperature range, including criticality.

Allowable 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 5-6. This is in addition to other criteria which must be met before the reactor is made critical, as discussed in the followmg paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the

. criticality limit line shown in Figures 5-1,5-3 and 5-5 (for the specific heatup rate being utilized). The straight line portion of the enticality limit is at the muumum permissible temperature for the 2485 psig mservice 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 Codem as follows:

1.5Kw < Ku (11) where,

- Ki,,, is the stress intensity factor covered by membrane (pressure) stress, Ki.= 26.78 + 1.233 l e ** **** *l, T is the muumum permissible metal W..p&ure, and  ;

RTum is the metal .4.e.ce nil-ductility t ww re .

1 i

~ The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margm during actual power production as specified in Reference 2. The pressure-temperature limits for J core operation (except for low power physics tests) are that the reactor vessel must be at a temperature ]

equal to or lugher than the mmimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the muumum permissible temp-rature in the corresponding pressure-temperature curve

- for heatup.and cooldown calculated as descriM in Section 3 of this report. The muumum temperature for  ;

the inservice hydrostatic leak test for the Byron Unit 2 reactor vessel at 16 EFPY is 227'F at 2485 psig ,

using the 1989 App. G Methodology,219'F at 2485 psig using the 1996 App. G Methodology and 166*F l l

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

' [(

i

5-2 at 2485 psig using Code Case N-588. 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.

Figures 5-1 through 5-6 defme all of the above limits for ensurmg prevention of nonductile failure for the Byron Unit 2 reactor vessel. The data points for the heatup and cooldown pressure-temperature limit curves show in Figures 5-1 through 5-6 are presented in Tables 5-1 through 5-6, respectively.

Additionally, Westinghouse Engineering has reviewed the muumum boltup temperature requirements for the Byron Unit 2 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 fothe material stressed by the boltup..

Therefore, since the most limiting initial RTer value is 30 F (vessel flange), the reactor vessel can be bolted up at this temperature.

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

53 MAhERIAL PROPERTY BASIS LIMITING MATERIALi CIRCUMFERENTIAL WELD WF-447 LIMITING ART VALUES AT 16 EFPY: 1/4T,94*F 3/4T,77'F 2500 , , ,, ., ,

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FIGURE 5-1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100*F/hr)

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

i Heatup and Cookkmn Pressure-Temperanire Limit Curves Revision 0 1-1 I

5-4 1

MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 -

LIMITING ARTVALUES AT 16 EFPY; l/4T,94'F - l 3/4T, 77'F

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FIGURE 5-2 Byron Unit 2 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 Margas forInstrunwatatimi Errors)

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

p.-

5-5 ks l .. TABLE S-1 Byron Unit 2 Heatup Data at 16 EFPY Using 1989 App. G Methodology (Without Margins of for Instnunentation Errors)

Heatup Curves Con 6guration #: 769956482 100 Heatup Critical. Limit Leak Test Linut T P T P T P i 60 0 227 0 206 2000 60 575 227 575 227 2485 65 575 227 575

85 575 227 575 90 575 227 575 95 575 227 575 100 575 227 575 105 575 227 575 110 $75 227 576 ,

115 576 227 579 l 120 579 227 585 r

125 585 227 592 130 592 227 601 -

135 601 227 612 140 612 227 625 145 621 227 640 150 621 227 657 i 150 640 227 676 l 155 657 227 697 160. 676 227 720  ;

165 697 227 745 l 170 720 227 773 175 745 227 804 l 180 773 230 837 185 804 235 873 l 190 837 240 912 195 873 245 954 200 912 250 999 205 954 255 1048 210 999 260 1100 215 1048 265 1157 220 1100 270 1217 225 1157 -275 1283 230 1217 280 1353 235 1283 285 1428 240 - 1353 290 1509 245 1428 295 1595 250 1509 300 1687 255 1595 305 '1786 I

l Heatup and Cooldown Pressure Temperature Limit Curves Revtsion 0

5-6 TABLE 5-1 (Continued)

Byron Unit 2 Heatup Data at 16 EFPY Using 1989 App. G Methodology (Without Margins of for Instrumentation Errors)

Heatup Curves Configuration #: 769956482 100 Heatup Critical. Limit Leak Test Lumt T P T F T P 260 1687 310 1892 265 1786 315 2005 270 1892 320 2126 275 2005 325 2254 280 2126 330 2391 285 2254 290 2391

. l l

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

5-7 I

TABLE 5-2 Byron Unit 2 Cooldown Data at 16 EFPY Using 1989 App. G Methodology

- (Without Margins of for Instrumentation Errors)

Cooldown Curves Configuration #: 769956482 Steady State 25F 50F 100F l T P T P T P T P 60 0 60 0 60 0 60 0 60 583 60 537 60 491 60 395 l 65 593 65- 548 65 502 65 408 70 603 70 559 70 514 70 422 .

75 615 75 571 75 526 75 437 )

80 621 80 584 80 540 80 453 I 85 621 85 597 85 555 85 470 90 621 90 612 90 571 90 489 95 621 95 621 95 588 95 510 100 621 100 621 100 607 100 532 105 621 105. 621 105 621 105 556 110 621 110 621 110 621 110 582 115 621 115 621 115 621 115 609 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 150 928 150 910 150 895 150 876 155 %3 155 948 155 937 155 927 160 1001 160 990 160 982 160 982 165 1042 165 1034 165 1031 165 .1041 170 1086 170 1082 170 1083 175 1133 18') 1183 j 185 1238 l

190 12 %

195 200 1358  !

1425 l 205- 1497 210 1575 215 1657 220 1746 225 1841 230 1942 235 2051 240 2167 245 2291 l 250 2423 Heatup and Cooldown Pressure Temperature Limit Curves Revision 0

V .

S-8 ]

1

, 1 MATERIAL PROPERTY BASIS LIMITTNG MATERIAL: CIRCUMFERENTIAL WG,D WF-447

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FIGURE 5-3 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100 F/hr)

Applicable to 16 EFPY Using 1996 Appendix G Methodology (Without Margins of for Instrumentation Errors)

Heanup and Cooldown Pressure-Temperature Linnit Curves Revision 0

-3 I

1 5-9 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 i 1/4T,94*F l LIMITING ART VALUES AT 16 EFPY:

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FIGURE 5-4 Byron Unit 2 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 Margins of for Instrumentation Errors) l 1

Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

5-10 TABLE 5-3 Byran Unit 2 Heatup Data at 16 EFPY Using 1996 App. G Methodology (Without Margins of for Instrumentation Errors)

Heatup Curves Configuration #: 1795436144 100 Heatup Critical. Limit Leak Test Limit T P T P T P 60 0 219 0 198 2000 60 621 219 635 219 2485 65 621 219 674 85 621 219 660 90- 621 219 650 95 621 219 643 100 621 219 638 105 621 219 637 110 621 219 637 115 621 219 641 120 621 219 646 125 621 219 654 130 621 219 664 135 621 219 676 140 621 219 690 145 621 219 707 150 621 219 725 150 707 219 746 155 725' 219 770 160 746 219 796 165 770 219 824 170 7% 220 855 175 824 225 889 180 855 230 926 185 889 235 %6 190 926 240 1010

  • 195 966 245 1057 200 1010 250 1107 205 1057 255 1162 210- 1107 260 1221 215 1162 265 1285 220 1221 270 1353 225 1285 275 1427 230 1353 280 1506 235 1427 285 1591 240 1506 290 1683 245 1591 295 1781 250 1683 300 1887 255 1781 305 2001 260 1887 310 2123 265 2001 315 2254 270 2123 320 2395 275 2254 280 2395 Heatup and Cooldown Pressure Temperature Limit Curves Revtsson 0

r 5-11 TABLE 5-4

( Byron Unit 2 Cooldown Data at 16 EFPY Using 1996 App. G Methodology (Without Margins of for Instrumentation Errors) l l

Cooldown Curves Configuration #: 1795436144 Steady State 25F 50F 100F T P T P T P T P 60 0 60 0 60 0 60 0 l 60 621 60 574 60 523 60 418 65 621 65 585 65 534 65 431 70 621 70 597 70 547 70 446 75 621 75 610 75 561 75 462 80 621 80 621 80 576 80 480 85 621 85 621 85 592 85 498 90 621 90 621 90 609 90 519 95 621 95 621 95- 621 95 541 100 621 100 621 100 521 100 564 105 621 105 621 105 621 105 590 110 621 110 621 110 621 110 618 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 150 995 150 974 150 957 150 935 155 1033 155 1016 155 1002 155 989 160 1075 160 1061 160 1051 160 1048 165 1119 165 1109 165 1104 165 1112 170 1166 170 1161 170 1161 175 1217 175 1217 180 1272 l 185 1331 l 190 1395 195 1463 200 1536 205 1615 210 1700 l 1

215 1791 220 1889 l 225 1995 l 230 2108 235 2230 240 2361 l.

l Heatup and Cooldown Pressure-Temperature Limit Curves Revision 0

w-5-12 1

MATERIAL PROPERTY BASIS  !

i LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING LIMITING ART VALUES AT 16 EFPY: 1/4T,94'F (N-588) & 4l'F ('96 App. G)  !

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FIGURE 5-5 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heaup Rate of 100*F/hr)

Applicable to 16 EFPY Using Code Case N 588 vs.1996 Appendix G with Axial ART (Without Margins of for Instrumentation Errors)

Heatup and Cooldown Pressure Temperature Limit Curves Revision 0

{ 5-13 I- MATERIAL PROPERTY BASIS L

1 LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING LIMITING ARTVALUES AT 16 EFPY; l/4T,94*F (N 588) & 4l'F ('96 App. G) -

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i FIGURE 5-6 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100*F/hr) Applicable to 16 EFPY Code Case N 588 vs.1996 Appendix G with Axial ART (Without Margms of for Instnamentation Errors)

[  !

1 Heatup and Cooldown Pressure-Temperature Limit Cuives Revision 0 i

5-14 TABLE 5-5 Byron Unit 2 Heatup Data at 16 EFPY Using Code Case N-388 vs.1996 App. G Methodology (Without Margins of for Instrumentation Errors)

Heatup Curves Configuration #: 361907406 100 Heatup Critical. Limit Leak Test Lmut T P T P T P 60 0 166 0 145 2000 60 621 166 620 166 2485 65 621 166 620 85 621 166 620 90 621 166 620 95 621 166 620 100 621 166 620 105 621 166 620 110 621 166 620 115 621 166 620 120 621 166 620 125 621 170 620 130 621 175 620 135 621 180 620 140 621 185 620 145 621 190 620 150 621 190 1080 150 1080 '195 1122 155 1122 200 1168 160 1168 205 1218 165 1218 210 1273 170 1273 215 1333 175 1333 220 1398 180 1398 225 1469 185 1469 230 1545 190 1545 235 1628 195 1628 240 1716 200 1716 245 1812 205 1812 250 1916 210 1916 255 2027 215 2027 260 2147 220 2147 265 2276 225 2276 270 2415 230 2415 Note: The computer run for the '% App. G using the highest Axial Flaw ART value generated the most conservative curve overall.

Hestup and Cooldown Pressure-Temperature Limit Curves Revision 0 i

i 5-15 TABLE 5-6 Byron Unit 2 Cooldown Data at 16 EFPY Using Code Case N-588 vs.1996 App. G Methodology (Without Margins of for Instrumentation Errors)

Cooldown Curves Configuration #: 361907406 l Steady State 25F 50F 100F T P T P T P T P 60 0 60 0 60 0 60 0 60 621 60 621 60 621 60 621 65 621 65 621 65 621 65 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 120 621 120 621 125 621 130 621 135 621 140 621 145 621 150 621 150 1583 155 1666 160 1754 165 1849 170 1952 175 2062 180 2180 185 2308 190 2444 Note: The computer nm for the '96 App. G using the highest Axial Flaw ART value generated the most conservative curve overall.

Heatup and Cooldown Pressure-Temperature Limit Cun'es Revtsion 0

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i

.6 REFERENCES i

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

l Nuclear Regulatory Commission, May,1988.

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

No. 243, dated December 19,1995.

3 1989 ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix Q " Fracture  ;

Toughness Criteria for Prbtection Agamst Failure". l l

4 CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI l Consulting,- March 1996. i 1

5 WCAP-14824, Revision 2, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal I

Operation and Surveillance Weld MetalIntegration For Byron and Braidwood", T. J. Laubham, et al., November 1997. Ref. Errata letter CAE-97-233, CCE-97-316, " Transmittal of Updated Tables to WCAP-14824 Rev. 2"

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

7 WCAP-15176, " Analysis of Capsule X from the Commonwealth Edison Co. Byron Unit 2 Reactor i Vessel Radiation Surveillance Program', T. J. Laubham, et al., March 1999.

8 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.

9. Letters CAE-97-231, CCE-97-314, " Comed Panaaaaa 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. BWwk & Wdcox drawing number 185265E, Revision 2; " Reactor Vessel General Outline
11. ASME Boiler and Pressure Vess:1 Code,Section XI, " Rule for Inservice Inspection of Nuclear Power Plant Cr--g-:=4", Appaa4 Q " 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 Asnect of Fracture Mechanics in Presars Va==-1c and Pinina- ed. S.S. Palusamy and S.G Sampath, PVP-Volume 58, ASME 1982.

6 References Revision 0

6-2

13. NRC SER Dated January 21,1998, " Byron Station, Units 1 and 2 and Braidwood Station. Units 1 and 2, Acceptance For Referencing 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 Regardmg Reactor pressure Vessel", From R.M. Krich, Dated September 3,1998.
15. NDIT No. MSD-98-044, "Best Estunate Chemistry Values for Reactor Pressure Vessel Beltline Weld Heat Number 442002", Dated December 1998.
16. WCAP-14940, " Byron Unit 2 Heatup and Cooldown Limit Curves For Normal Operation", T. J.

Laubham, October 1997. Ref. Errata letters CAE-97-210 & 232, CCE-97-289 & 316, 1

" Transmittal of Updated Tables to WCAP-14940 and WCAP-14970" l i~ WCAP-15180, " Commonwealth Edison Company Byron Unit 2 Surveillance Program Credibility Evaluation", T.J. Iaubham, March 1999.

18. WCAP-15123, Revision 1, " Analysis of Capsule W from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", T.J. Laubham, et. al., January, 1999.

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References Revision 0

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