ML20199E828

Heatup & Cooldown Limit Curves for Normal Operation

ML20199E828
Person / Time
Site:	Byron
Issue date:	10/31/1997
From:	Christopher Boyd, Howell D, Laubham T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20199E810	List: ML20199E806 ML20199E828 ML20199E853 ML20199E875 ML20199E880
References
WCAP-14940, NUDOCS 9711240009
Download: ML20199E828 (33)

Text

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(, .i BYRON UNIT 2 HEITEP -

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CURVES FOR NORMAL OPERATION

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Byron Unit 2 Heatup and Cooldown Limit Curves For Normal Operation T. J. Laubham October 1997 Work Performed Under Shop Order CPEP-139 Prepared by the Westinghouse Electric Corporation for the Commonwealth Edison Company Approved: .

~C . H . Boyd, Manaher M

Equipment & Msterials Technology Approved:

D. A . Howell, Manager Mechanical Systems integration WESTINGHOUSE ELECTRIC CORPORATION Nuclear Services Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1997 Westinghouse Electric Corporaition All Rights Reserved 9

._. _ _ , . , _ , . --m- v... + - . . . , . . . _ . - , _ . - _ _ - - . - . .

PREFACE. .

i TW report has been technically reviewed and verified by:,

E. Terek~

r-r

- Byron Unit 2 Hertup and Cooldown Limit Curves October 1997

An. u TABLE OF CONTENTS .

Li ST OF FlG U R E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 f ,

LI ST OF TAB LE S . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1 I NTR OD U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . .

(

2 2 FRACTURE TOUG HNESS PR O"ERTIES...............................................................

1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS.. 3 8

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

5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES........16 6 R E FE RE N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

lii LIST OF FIGURES 1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100'F/hr) Applicable for the First 12 EFPY (Without Margins for i nstrume ntr. tion Errors) . .. . . . . . . . . . . . . . . . . . .. .. . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . .... . . . . . . . . . . . . .. . . . . . . . . . . . . .. . . . . 18 2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 12 EFPY (Without Margins for i nstrumentation Ermrs ) . .. . . . . . . . . . . . . . . . .. . . . . . . . .. . . . . ... . ... . . . .. . . .. . .. . .. . . . . . . . . . . .. . . . .. . . . . . . . . . . . . . . . . . . ... 19 s

3 Byron Unit 2 Reactor Coolant System Heatup Limitations (Hestup Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for }

instrumentation Errors; Using 1996 Appendix G Methodology)............................... 20 ,4; 4 Byron Unit 2 Reactor Loolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Matgins for instrumentation Errors: Using 1996 Appendix G Methodology)................... ........... 21 C

6 Byron Unit 2 Heatup and Cooldown Umit Curves October 1997

iv LIST OF TABLES 1 Calculation of Average Cu and Ni Weight % Vataos for the Byron Unit 2 _

Bas 3 Metals..............................................................................................................9 2 Calculation of Average Cu and Ni Weight % Values for the Eyron Unit 2 Weid unterial (Using Byron 1 & 2 Chemistry Test Results)..................................... 10 Byron Unit 2 Reactor Vessel Matenal %+"J n..................................................... 10 3

4 Calculation of Chemistry Factors Using Credible Byron UrAs 1 and 2 Surveillance Capsule Data . .. . . . . . . .. . ... .. .. . .. . . . . .. .. . . . .. . .. ... ... . . .. .... .. . . . . . . .. . .. . . .. . . .. . . .. . . ... . . . . . . 11 5_ Calculebon of Adjusted Reference Temperatures (ART) at 12 EFPY for all Byron Unit 2 Reactor Vessel Material (tissed on ommene survesence capsule date)........... 14 6-- Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T Locations for 12 EFPY. . .. . . .. .. . . . . . . .. . . . . . . . .... . . .. . . . ... . .... . .. . . . .. . . .. . .. . ... . . ... . . ... ..... . . .. . . . . . .. . . .. . . 15 __

7 Byron Unit 2 Heatup Data at 12 EFPY Wlthout Margits for instrumentation ,

Errors (inoiueen veneel new mouwements of 1WF end 621 pois por 10CFR50), ................... 22 8 Byron Unit 2 Cooldown Data at 12 EFPY Without Margins for Instrumentabon Errors (includes veneet aange requwements of 1FF and 621 peig por 10CFR50). .................. 23 9 Byron Unit 2 Hestup Data r.t 12 EFPY Without Margms for instrumentation Errors, Us6ng the 1996 App. G Methodology (Inciudes veneet aange requwements of 150T and 621 peig per 10CFR50)................................ 24 10 Byron Unit 2 Cooldown Data at 12 EFPY Without Margins for instrumentation Errors, Using the 1996 App. G Methodology (ineeudes veneel aange requwements of 1WF and 621 poig per 10CFR60)................................ 25 e

Byron Unit 2 Hestup and Cooldown Umst Curves October 1997

1

-lNTRODUCTION

-1 Herup and cooldown limit curves are calculated using the adjusted RT,et (reference nil-ductility tmperature) corresponding to the limiting beltline region material of the reactor vessel. The adjtsted RT,er of the limiting materialin the core region of the reactor vesselis determined by using the unstradated reactor vessel material fracture toughness properties, estimating the radetson-induced ART,et, and adding a margin. The unitradiated RT, y is designated as the higher of either the drop weight nil ductility tran*%n temperature (NDTT) or the temperature at which the matenal eichibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.

RT,er increases as the material is exposed to fast-neutron redation. Therefore, to find the most limiting RT,er at any time period in the reactor's life, ART,e, due to the radiation exposure associated with that time period must be addod to the unirradiated RT,er(IRT,er). The extent of the shift in RT,ev is enhanced by certain chemical elements (such as copper and nickel) present

.in reactor vessel stoeis. The Nuclear Regulatory Commission (NRC) has published a method for predicting redation embrittlement in Regulatory Guide 1.gg, Revision 2, " Radiation Embnttlement of Reactor Vessel Materials *3. Regulatory Guide 1.gg, Revision 2, is used for ,

the calculation of Adjusted Reference Temperature (ART) values (IRT,et + ART, y + margins '

, far uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the

, beltline region measured from the clad / base metalinterface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves.

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

2,,

\

2- FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the 'emtic material in the reactor coolant pressure m

- boundary are determined in accordance with the NRC Regulatory Standard Review Plan . The pre-irradiation fracture-toughness properties of the Byron Unit 2 reactor vessel are presented in Table 3. Credible surveillance data is available for two capsules (Capsules U and W) for Byron Unit 2. The post-irradiation fracture toughness properties of the reactor vessel surveillance material was obtained directly from the Byron Unit 2 Reactor Vessel Radiation Surveillance Program Results'*3 and was used to calculate chemistry factors (See Table 4). For all other beltline materials the chemistry factor was calculated per Regulatory Guide 1.99, Revision 2, l position 1.1.

Per the request of the Commonwealth Edison Company, the surveillance weld data from the Byron Unit 1 and Byron Unit 2 surveillance programs" *1 has been integrated in addition to the credible surveillance weld data from Byron Unit 2, credible surveillance weld data is available for two capsules (Capsules U and X) for Byron Unit 1. The chemistry factor values resulting from the weki metalintegration of the Byron Unit 1 and 2 surveillance program results is presented in Table 4 in Section 4 of this report.

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

3 s

3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements'" specifies fracture toughness requirements for ferritic materials of preseure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety durir.g any condition of normal operation, including anticipated operational occurrences 4

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 4'1, V;ssels, contain the conservative methods of analysis.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K,, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K , for the metal temperature at that time. K., is obtained from the reference fracture toughness curva, defined in Appendix G of the ASME Code,Section XI. The K, curve b given by the following equation:

L = 26.78 + 1.23 3t

• c """'"'"""" (1) where, K,, = reference stress intensity fac. tor as a function of the metal temperature T and the metal reference nil-ductility temperature RT.

Therefore, the goveming equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C' K n,+ L < L (2) where, K ,, = stress intensity factor caused by membrane (pressure) stress K, = stress intensity factor caused by the thermal gradients K., = function of temperature relative to the RTem of the material C= 2.0 for Level A and Level B service limits C= 1.5 for hydrostatic and leak test conditions during which the reactor core is not cri+Jeal Byron Unit 2 Heatup and Cooldown Limit Curves October 1997 E _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - -

m y.

~

c g

<=

j

At cny time during the hentup cr cooldown tr:nsienti the cllow ble v lue of N b det:rmined by.

tM metal temperature at the tip of a postulated fisw at the 1/4T and 3/4T location, the -

oppropriate value for RT,m, and the reference fracture toughness curve.' The thermal stresses -

resulting from the temperature gradients through the vessel wall are calculated and then the correspondmg (thermal) stress intensity factors, K., for the reference flow are computed. From

= Equat6on 2, the pressure stress intensity factors are obtained and, from these, the allowable -

pressures us calculated. ,

For the ceksulation of the allowable pressure versus coolant temperature dunng cooldown, the reference fisw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel-wall. Dunng coolr/own, the controlling location of the flew is always at the inside of the wall because the thermal gradier ts produce tensile stresses at the inside, which increase with-increasing cooldown rates. Allowable pressure-temperature relations are generated for both ~

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

. The use of the cornposite curve in the moldown analysis is necost.ary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed I -

flow. During cooldown, the il4T vessel locahon is at a higher temperature than the fluid ,

' ad}ecent to the vessel inner diameter. :This condition, of course, is not true for the steady-state

- situation. It follows that, at any gWen reactor coolant temperature, the AT (temperatuni.)

developed during cooldown results in a higher allowable value of N st the 1/4T location for finita cooldown rates than for steady-state operation. Furthermore, if condebons exist so that the increase in allowable value of N exceeds K., the calculated allowable pressure dunng cooldown will be greater than the steady-state value. 3 The above procedures are needed because ther'r is no direct contro! c. temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of coolmg is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. ,

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by intomal pressure. The metal temperature at the crack tip legs the coolant temperature; therefore, the allowable value of N for the 1/4T crack during heatup is lower than the allowable value of N for the 1/4T crack during -

tteady-state conditions at the same coolant temperature. During heatup, espscially at the end cf the transient; conditions may exist so that the effects of compressive thermal stresses and lower allowable Nvalues do not offset each other, and the pressure-temperature curve based Byron Unit 2 Hestup and Cooloown Limit Curves Octobe* 1997

5 on steady state conditions no longer represents a lower bound of all simliar curves for finite heatop rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the u!!owable pressure calculated for steady state and finite heatup rates is obtained.

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 b assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the bestup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each bestup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady state and finite heatcp rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken fr m 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 RT,ev by at least 120*F for normat operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is 621 psig for Byron Unit 2.

The limiting unitradiated RT,e1 of 30*F occurs in the vessel flange of the Byron Unit 2 reactor vessel, so the minimum allowable temperature of this region is 150*F at pressures greater than 621 psig. This limit is shown in Figures 1 through 4 wherever applicable.

1996 Addenda to ASME Section XI, Appendix G MethodologyN Appendix G was recently revised to incorporate the most recent elastic solutions for Ki due to pre:sure 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 NI;O, 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 ar.:t Newman"M.

This revision now provides consistent computational methods for pressure and thermal K, for Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

6 l

thermal gradients through the vessel w:ll ct Cny time during the transient. Consist:nt with tha original version of Appendix G, no contribution for crack face pressure is included in the F Ne 13 pressure, and cladding effects are neglected.

Using these most recent elastic solutions in the low temperature region will provide some retret t) restrictions associated with reactor operation at relatively low temperatures. Although the relief b relatively smallin terms of absolute allowable pressure, the benefits are substantial because even a smallincrease in the allowable pressure can be a significant percentage increase in the operating window at relatively low temperatures. Implementing this revision results in an economic and potential safety benefit (less likelihood of lifting 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 2 at 60*F is 25 psig.

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 E < 2, M,,, = 0.9268 for 25 4 5 3.464, M,,, = 3.21 for E > 3.464 Similarly, M,,, for an outside surface flaw is given by:

M,,, = 1,77 for E < 2, M,,, = 0.8934 for 2s 8 5 3.464, M,,, = 3.09 for E > 3.464 and p = intemal pressure. Ri = vessel inner radius, and t = vessel wall thickness.

G-2214.3 Radial Thermal Gradient:

The maximum K, produced by radial thermal gradient for the postulated inside surface defect of G-2120 is K = 0.953x104x CR x t 25 , where CR is the cooldown rate in 'F/hr., or for a postulated outside surface defect, K = 0.753x104x HU x t 5, 2 where HU is the heatup rate in

'F/hr. >

The through-wall temperature difference associated with the maximum therma! K, can be determined from Fig. G-2214-1. The temperature at any radial distance fro:n the vessel surface can be determined from Fig. G-2214-2 for the maximum thermal K, .

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

7 The maximum thermal K, relationship and the temperature relationship in Fig. G-2214-1 (a) are applicable only for the conditions given in G-2214.3(a)(1) and (2). 4 (b)_

Attematively, the K, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a %-thickness inside surface defect using the relationship:

L = (1.0359Co + 0.6322Ci + 0.4753C + 0.3855Cs)

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

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

• 6 (5) where the coefficients Co, C,, C, and 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) + Ca(x / a)2 + C3(x / a)' (6) and x is a varible 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.

Byron Untt 2 Heatup and Cootdown Umit Curves October 1997 l

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

ART = JnitialRTwor + b RTpor + Margin (7)

Initial RT, is the reference temperature for the unirradiated material as defined in paragraph f

NB.2331 of Section 111 of the ASME Boiler and Pressure Vessel Code't. If measured values iniel RTc for the material in question are nc uvallable, generic mean values for that class of material may be used if there are sufficient test resuRs to establish a mean and standard deviation for the class.

ART,c,is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

A RTsor = CF *f*"* zey To calculate ARTc taany depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence to the specific depth. ,

f,w =f *c"* (9) where x inches (vessel beltline' thickness is 8.5 inches"81) 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 ART,ev at the specific depth. The calculated surface fluence for Byron Unit 2 upper and lower shell forgings and circumferential weld at 12 EFPY is 8.22 x 10" n/cm'. This fluence value was calculated from the surveillance Capsule W analysis presented in WCAP-140640 'l >

CF ('F) is the chemistry factor, obtained from the tables in Reference 1, using the average values'of copper and nickel content as calculatea in Table 1 and 2, and reported in Table 3.

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

The Ratio Procedure, as documented in Regulatory Guide 1.99 Revision 2 Position 2.1, was used to adjust the measured values of ART ex of the weld metal, for differences in copper and nickel content, by multiplying them by the ratio of the chemistry factors for the vessel material (best-estimate chemistry) to that for the surveillance weld.

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997 E --

9 All materials in the beltline region of the Byron Unit 2 reactor vessel were considered in I determining the limiting material. The calculations to determine the ART values for beltline materials et 12 EFPY are shown in Table 5. The resulting ART values for all beltline region materials at the 1/4T and 3/4T locations are summarized in Table 6, where it can be seen that the limiting material is the circumferential weld (based on credible surveillance capsule data),

The 1/4T and 3/4T ART values for circumferential weld (based on credible integrated surveillance capsule data) were used in the generation of heatup and cooldown curves applicable to 12 EFPY, TABLE 1 Calculation of Average Cu and NiWeight % Values for the Byron Unit 2 Base Metals inter. Shell Forging Lower Shell Forging 49D329/49C297 1 MK 24-2 49D330/49C298 1 MK 24 3 Reference Cu % Ni% Cu% Ni%

WCAP -14063 0.01 0.70 0.07 0.65 0.006- 0.70 0.022 0.689 Byron Unit 2 0.006 0.71 0.05 0.73

~

Hectup & Cooldown 0.05 0.73

~

for Normal 0.06 0.75 Operation V83 0.067 0.772 Average 0.01 0.70 0.05 0.72 Standard Deviation 0.002 0.01 0.02 0.04

\

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997 I

10 I

l TABLE 2 Calculation of Average Cu and Ni Weight Percent Values for the Byron Unit 2 Weld Matenal (Using Byron 1 & 2 Chemistry Test Results)*

Wold Type Cu% Ni% -

B&W Wold Qualificatons 0.024 0.70 (BAW-2261) 0.031 0.46 0.03 0.72 0.068 0.48 0.114 0.54 0.148 0.60 0.053 0.6?.

0.059 0.62 B&W Wold Qualificaten* 0.029 0.65 Byron 1 Sury. Wold Data

• 0.022 0.690 Byron 2 Sury. Wold Data

• 0.023 0.712 Best Estimate Chemistry 0.055 0.617 Standard Deviaton 0.042 0.091 NOTES:

(a) The weld matenalin the Byron Unit i surveillance program was made of the sarra wire and flux as the reactor vessel intermedete to lower shell girth seam weld. (Weld seam \W-336, Wire Heat No. 442002, Flux Type Linde 80, Flux Lot No. 8873)

(b) The Byron Unit 2 surveillance weld is identcal 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, not number 8064.

Cu & Ni values obtained from WCAP-14824 Rev. U") i (c) I (d) Obtained from Reference 17.

TABLE 3 Byron Unit 2 Reactor Vessel Material Properties Matenal Desenption CU (%) Ni(%) Chemistry initial i Factor

• RT,ey (*F)* .

Closure Head Flange Not Reported 0.74 - 0'S Vessel Flange Not Reported 0.73 - 30'*

Intermediate Shell Forging 0.01 0.70 20 -20 49D329/49C297 -1 1

~

Lower Shell Forging 0.05 0.72 32.2 -20 490330/49C298 1 Circumfetential Weld 0.05 0.62 68.0 10 NOTES:

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

(b) Inital RT,e, values are measured values.

m (c) Closure head and vessel flange inliial RT,e, values are used for considenng flange requirements for the heatuplcooldown curves.

Byton Unit 2 Heatup and Cooldown Lim.t

• Curves October 1997

11 TABLE 4 Calculation of Chemistry Factors Using Credible Byron Units 1 and 2 Surveillance Capsule Data Material Capsule Capsule FPS Measured FF* FF' Fluence f ART,e7 ART,e7 Lower Shell Forging 49D330/ U 3.996 x 10 O.746 0 0 0.556 49C298-1 1 (Tangental)

W 1.211 x 10 1.053 5 5.267 1.110 Lower Shell Forging U 3.996 x 10 O.746 25 18.65 0.556 .

49D330/49C298-1 1 (4x,i) W 1.211 x 10 1.053 40 42.12 1.110 Sum: 66.037 3.332 Chemistry Factor'* = 66.046 + 3.332 = 19.8'F Byron 1 Weld U 3.72 x 10 O.727 0 0 0.00 0.529 MetalWF 336*)

X 1.39 x 10 1.091 35 105'" 114.56 1.19 Byron 2 M U ' 3.996 x 10 O746 0 0 0.00 0.557 Metal WF 447'4 W 1.211 x 10 1.053 30 90'* 94.77 1.110 Sum: 209.33 3.386 Chemistry Factor'* = 209.33 + 3.386 = 61.8'F NOTES:

() FF = Fluence Factor = f8 * "***

(b) Byron Unit 1 ART,e7 values were obtained from the surveillance Capsule X analysis (WCAP-13880).

The Byron Unit 1 capsule fluence values were recs,1culated using the ENDF/B V scattenng cross sectons in 1994 and are documented in WCAP-1404481 (c) Byron Unit 2 capsule fluence FF, and ART,o1 values were obtained from the surveillance Capsule W analysis (WCAP-14064"'l) using the ENDF/B-V scattenng cross sections.

(d) Chemistry Factor = E(FF* ART,ey) + I(FF2)

(0) Adjusted ART,e7 per Rato Procedure of Regulatory Guide 1.99 Rev. 2. Ratio = 3.0. Actual ratio is 2.5 (68.0 + 27.0 = 2.5), however, for conservatism a ratio of 3.0 was used in this case.

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

12 Explanation of Margin Terms used form _ Byron Unit 2 When there are 'two or more credible surveillance data sets"N available for Byron Unit 2, Regulatory Guide 1.99 Rev. 2 (RG1.99R2) Position 2.1 states "To calculate the Mar 9i n in this case, use Equation 4, the values given there for og may be cut in half" Equation 4 from R31.99R2 is as fcilows: M = 2da? + al .

Standard Deviation for initial RT,e7 Margin Term, o, if the initial RT,c, values are measured values, which they are in the caoe of Byron Unit 2, then eib equal to O'F, On the other hand, if the initial rte values were not measured, then a generic value of 17'F would have been required to be used for o,,

Standard Deviation for ART,e1 Margin Term, og Per RG1.99R2 Position 1.1, the values of og are referred to as "28'F for welds and 17'F for base metal, except that og need not exceed 0.50 times the m6an value of ART,er." The mean value of ART,ev is defined in RG1.99R2 by Equation 2 and defined herein by Equation 8.

Per RG1.99R2 Position 2.1, when there is credible surveillance data, og is taken to be the lesser of %' ART or 14*F (28'F/2) for welds, er 8.5'F (17'F/2) for bass metal. Where ART,e1 ogsin is defined herein by Equation 8. (

Summary of the Margin Term Since o,is taken to be zero when a heat-specific measured value of initial RT,e1 are available

- (c.s they are in th',3 case), the total margin term, based on Equation 4 of RG1.99R2, will be as

. follows:

o Position 1.1: Lesser of ART,e1 or 56*F for Welds Lesser of ART,e1 or 34*F for Base Metal o Position 2.1: Lesser of ART,e1 or 28'F for Welds Lesser of ART,e1 or 17'F for Base Metal Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

13 The following is a sample calculation of the margin t:rm for the weld metal at the % T location. .,

The results for this calculation as well as the results for the remaining reactor vessel beltline materials are documented in Table 5.

Margin Term for Wold Metal (1/4T Location):

o From Equation 8

• ART,cv = CF x FF whens, CF = 68.0 (R.G. Position 1.1)

= 61.8 (R.G. Position 2.1; i.e. using Sury. Caps. Data)

FF = 0.803 (@ 12 EFPY and Fluence = 8.22 y ~ T' n/cm')

Therefore, ARTet = 54.60 (R.G. Position 1.1)

= 49.63 (R.G. Position 2.1; i.e. using Surv. Caps. Data) o From Equation 4 (of R.G.1.it9 R2)

• M = 2da' + a' where,  % A R T ,e v = 27.30 (R.G. Position 1.1)

= 24.815 (R.G. Position 2.1; i.e. using Surv. Caps. Data) e, = C'F (Initial RT,ev is Measured) o4 = Lesser of (% ART,et ) or (28'F)

= 27,30 (R.G. Position 1.1) og = Lesser of (% ART,cr ) or (14'F)

= 14.00 (R.G. Position 2.1; i.e. using Sury. Caps. Data) 2 8 Therefore, M = 2J0 + 273 = 54.60 (R.G. Position 1.1)

M = 240' + 14.0' = 28.00 (R.G. Position 2.1; i.e. usir;g Surv.

Caps. Data)

Byron Unit 2 Heatup and Cooldown Umst Curves October 1997

g - -

~

TABLE 5 Calculation of Adjusted Reference Temperatures (ART) at 12 EFPY for all Byron Unit 2 Reactor Vessel Material (based on credible surveillance capsule data)

Reactor VesselBenline Malenal f @ 12 Hi% CF" EFPY Lt f* Wt FF 8 ART " o, o. M ART * .

Renonloca6an identilles6an Cu%

~ 8 Lt FF

. (x 10 ) Etf

%T Calculalon 490329-11 0.01 010 20.0 0 822 0.494 0.303 -20 16.06 0 8.03 l 16 06 12.1 intennediate Shet Forging 49C297-1 0 822 0.494 0 803 -20 25 86 0 12.93 25.86 313 Lower Shes Forgmg 490330-11 0.05 0.72 32.2 i 49C298-1 0.494 0 803 -20 15.90 0 7.95 15.90 11.8 Lower shew Forging 19 8 0.822

-o using GfC Data l 0.822 0.494 0.803 10 54.60 0 27.30 54.60 119.2

! Weld Metal WF-447 0.05 0.62 68.0 (Heal 442002) 0.494 0.803 10 49.63 0 14.00 28.00 87.6 Weld Metal 61.8 0.822

-o using S/C Data l

%T Calcuhdon 0.178 0.542 -20 10.84 0 5.42 10.84 11 Intermediate Shel Forging 490329-11 0.01 010 20.0 0.822 49C297-1 0.178 0.542 -20 17.45 0 8125 17.45 14 9 Lower Shen Forging 490330-11 0.05 012 32.2 0.822 4NI 4 19.8 0.822 0.178 0.542 -20 1013 G 5.375 1013 _. __ 1.5 l Lower shell Forging

-+ using S/C Data 0.62 68.0 0.822 0.178 0.542 10 36.86 0 18.43 36.86 831 Weld Metal WF-447 0.05

..iuear442002L 61.8 0.822 0.178 0.542 10 33.50 0 14.00 2800_ 71.5 Weid Metal

-o using S/C Data NOTES:

(a) The Byrcn Unit 2 reactor vessel war thickness is 8.5 inches at the beltline region (b) ART = I + ART,cr + M (This value was rounded per ASTM E29. using the " Rounding Method".)

(c) ART,or = CF

• FF (d) The CF is integrated between the Byron 1 Weld (WF-336. heet # 442002) and the Byron 2 Weld (WF-447. Heat # 442002).

O d46 1997 Byton Unit 2 Heatup and Cooldown Lirnit Curves

15 1

TABLE 6 Summary of Adjusted Reference Temperaturis (ART) at 1/4T and 3/4T Locations for 12 EFP f Matenal 12 EFPY 1/4T ART 3/4T ART '

intermediate Shell Forging 12.1 1.7 49D329/49C297 1 (RG Posibon 1(a))

using credible surveillance 31.7 14.9 -

capsule data (RG Posthon 2(a))

Lower Shell Forging 11.8 1.5 490330/49C298 1 (RG Position 1(a))

Circumferential Weld 131.9 98.9 (RG Posinon 1(a))

using credible surveillance 87.6*) 71.5*)

capsule data (RG Position 2(a))

NOTES:

(a) Calculated using a chemistry factor based on Reguictory Guide (RG) 1.99, Revision 2, Posibons 1 and 2.

(b) These ART values were used to generate the B3mn Unit 2 heatup and cooldown curves.

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

16 i ~6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT LCURVES-Pressure-temperature limit curves for normal hoatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel belthne region using the methods"9 discussed in Section 3 and 4 of this report. The 1989 edihon methodology is also presented in WCAP 14040-NP-Al"3, dated January 1996.

Figures 1 and 3 present the heatup curves, using the 1989 and 1996 Appendix G Methodology respectively, without margins for instrumentaten errors and for a heatup rate of 100*F/hr appbcable for the first 12 EFPY. Figures 2 and 4 present the cooldown curves, using the 1989 ~

and 1996 Appendix G '.'-h-I-:ksy respectively, without margins for instrumentation errors and

.for cooldown rates up to 100*F/hr applicable for the first 12 EFPY. Allowable combinstions of #

temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 through 4. This is in addition to other criteris which must be met before the reactor is made enhcal. (As a note for Figures 1 through 4, the horizontal axis is the reactor water temperature or moderator temperature and the vertical axis is the calculated allowable pressure based on through wall stresses.)

- The reactor must not be made entical until pressure-temperature combinations are to the right of the enticality limit line shown in Figures 1 and 3. The straight-line portion of the criticality limit b at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The goveming equation for the hydrostatic test is defined in Appendix G to Section XI of the ASME Cuda as follows:

1.5Khi < Ku (10) where, K. is the stress intensity factor covered by membrane (pressure) stress,

- (= 26.78 + 1.233 e l***"* " *l, '

T is the minimum permissible metal temperature, and RT,e7 is the metal reference nil <$uctility temperature The criticality limit curve specifies pressure-temperature limits for core operation to provide addihonal margin during actual power production as specified in Reference 5. The pressure-temperature limits or core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and Byron Unit 2 Heatup and Coo 4down Limit Curves October 1997

~ ~ -

17 cooldown calculated as described in Section 3 of this report. The minimum temperature for the inservios hydrostatic leak tests for the Byron Unit 2 reactor vessel at 12 EFPY is 221'F @ 2485 psig using the 1989 App. G Methodology ~and 212*F @ 2485 psig using the 1996 App G Methodology. The verticalline 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 1 through 4 define all of the above limits for ensuring prevention of nonductile failure for the Byron Unit 2 reactor vessel. The data points used for 'he hestup and cooldown pressure-temperature limit curves shown in Figures 1 through 4 are presented in Tables 7 through 10.

Additionally, Westinghouse Engineering has reviewoo the minimum boltup temperatu e

- requirements for the Byron Unit 2 reactor pressure vessel. According to Paragraph G4222 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 lattial hydrostatic test pressure at the initial RT of the material stressed by the boltup. Therefore, since the most limiting initial RT value is 30*F (vessel flange), the reactor vessel can be bolted up st this temperature.

However, based on historical practices and engineering judgement, Westinghouse recommends a bolt up temperature of no less than 60*F.

i Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

l 1 18

~

e MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD (usmo sury.capsuk data)

LIMITING ART VALUES AT 12 EFPY: 1/4T, 87.6*F 3/4T, 71.5'F 2500 4 . , , ,

f  ;

,Y2250  ;  ;,

!i'ILEAE _JTEST LIWlf j y; r I I

[ i

[ ,

i i

l' i i 1 , 1-I I ii! i

" 2000 i 4 i

/ / / i e#

I I ii t i6 i ; i ;-

v i i i i j'

l I

I i

i i

I i

, . . I e-1750 l 6 UNACCEPTAaLE r r w '

OPERATION i i g f [

ua 1500 '

s r U3 ,

/ ,

/ ,

• / /

1250 '

f f I

g b

NEATUP RATE f

/ ACCEPTABLE UP TO 100 F/Er, f,

.r f

.r OPE, RATION i

' r1 r 1000  : '

/ '

m i e-

/ l  !':

e cu  !

, / / ._

750 i  :

m  ; ' ' ' l .  :

! i

,,,,, ,,  !, l, 2 500  :, ' ,4 ,,

i i-Q ,e i  : ,  ; , ;l

, N. , . , c

- i & i i i

' 's i I f e M

250 ' '

Bd"P '

i i !

i . i,i Temp.

()

I i 4 6 i jn j,jjj j j tj,g[gj j,' P,jj ag,gg TEMPERATURt 338 F i POR TBS

'g * '

0

,, SERTICE P t t l e( t UP PG 13.0 RFPY  ! '

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

FIGURE 1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for instrumentation Errors)

Byron Unit 2 Heatup and Cooldown Lirnit Curves October 1997

19 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD (using surv. capsule data)

LIMmNG ART VALUES AT 12 EFPY: 1/4T, 87.6'F 3/4T, 71.5'F 2500 _. z

_ ............. i , ,

-c., -

i >

- ' i

, +

me2250 l

l l  :' ;l l','

l

l l

l l

ll +-

i

• f ,

i . .

ur i ,

4 ,/

4 j- , i i  ! t I a a 2000 , ,

l . , '[

, 4 ,

l l++-

I I I .

, r--

I { i

e 1750

/ '

s t *1

___ UNACCEPTABLE v

^ '

y ___. 0PERATION I g _-- I i

/

, , 1500 '

m -..

/ p

, 1250 ,

/

T-L / AccrPTABLE _._

) OPERATION

/ . .

T 1000  ; , ,

/

i 4 i i '  !

" l cootnowM

• g, 750 -- -

• • "8 TrEr. .

j

~

2

o .

sm <

O 500 - -

ll too T,'

ei m 250  %

' y_

g ,

Boltup

,,, l l:l j

Temp. ,j '

i l

i'i ,

0 ';

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

FIGURE 2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 12 EFPY (Wdhout Margins for Instrumentation Errors) yron Unst 2 Heatup and Cooldown Limit Curves October 1997

20

~MATERIAL PROPERTY BASIS I

LIMITING MATERIAL: CIRCUMFERENTIAL WELD (using sury. capsule data)

LIMITING ART VALUES AT 12 EFPY: 1/4T, 87.6*F 3/4T, 71.5'F 2500 .

, , ,i

- s s essisue ,

l

!  ; j l ll,:

2250 I ' ' i i i !. .

ll V3

___ LEAE TEST LIMIT d;J ,

,I i

'I 6

I I  !

i , * * '

"*2000 I'i ' , i i

I I f i* i i i i i

, , l ,

Y.

f f

1 r ,i f 1

• o 1750  ; '

UNACCEPTABLE 1 / f i

6 OPERATION ,

l

o [ [ l m 1500 ,

l / /

,,i i i  ; 'l

@ i! i i 1

I

.s 4 I 4

/ 7 / t i C.) I i

, 1250  !' REATUP RATE Ur To soo r/s,. [ ,j j

li.

g

/ AccEPTAstE , , .

• 8

' i e

i i i 3

i ! / /

OPERATION i i -

6 i i r i 6 .' '

1000 l!  ! t (

-o i .

,. i

'l  !i O = - -

a 750 ,' '

m ' ,' '

m I i z 500 '

,c  ;,,

Q I i . ,

• %. ~I N i ,

~

i i i

'A

• 250 fiolmP ' y t

4 i

Temp.

!!. iiiii'Ir.i!!?.!!!'! !?

r u i. , , ,,,

' ' ' '  ::;;!!7!!ili'e,'4.';:.:'i,,,  !

0 ' '

i , , ,

0 50 100 150 200 250 300 350 400 450 Moderator Temperature (Deg.F) 500 FIGURE 3 Byron Unit 2 Reactor Cooiant System Heatup Limitations (Heatup Rates up to'100*F/hr) Applicable for the First 12 EFPY (Without Margins for instrumentation Errors: Using 1996 Appendix G Methodology)

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997 L .

21 l 4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD (using surv. capsule dets)

LIMITING ART VALUES AT 12 EFPY: 1/4T,87.6*F -

3/4T, 71.5'F 2500 ,

- si........ ,

4 i ,,. . .

^ '

w 2250 l

, 4

' l '! .i l l l l ,

l ,,; ,l l,'

.- 4 i i i > ,

, ., i .

m 2 2000 -

l l l' .l l

, j  ; ;l ,

lll ,l, ii i i , i

/ '

';' 'l 6 i 1e  ! i  ;

i , i o 1750 _._, UNACCEPTABLE [ '

y. +

__. OPERATION /

f m 1500 f m / ,

6 1250 '

ACCEPTABLE g / 0PERATION r

e 1000  ;

U -

C00LDOWN ce 750 -: -

005 ' I 0 j

- - - j -

i

• 500 -  : 0 "/ '

y [ 100 ,

ce 250 *-

u B InIP Temp.

'l

. , ;i 0 '!'' '  ! 'I  !

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

FIGURE 4 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 12 EFPY (Without Margins for instrumentation Errors; Using 1996 Appendix G Methodology)

Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

22 .

TABLE 7 Byron Unn 2 Hestup Data at 12 EFPY Without Margins for Instrumentation Errors (includes Vessel Flange requirements of 150*F and 621 psig per 10CFR50)

Heatup Curves (56assessws) 100 Hestup - 'Cntcal Limit ' Leak Test Limit >

.T P T P- T P

- 60 0 221 0 199 2000 60 594 ~221- 637 221 2485 65 594- 221- 623

. 70 594 221- 611:

- 75 594 221- 603 80 594 221 597 85 594 221 594 90 594 221 594 95 594 221 595

- 100 594 221 599 105 594 221 606 110 594 221. 614 115 ~ 595 221 624

' 120 599- 221 636 125 606 221 651 130 614 221 667 135 621 221 686 140 621 221- 707 145 621 221- 730 150 621 221 755 150- 667- 221 783 155 686 221 814 160- 707 225 847

- 165 730 . 230 E"3 170 755 235 921 175 783 240 964 180 814 245 1009 185 847 250 1058 190 883 255- 1111 l 195 921 260 1168 l 200 964 265 1229 l 4 205 1009 270 1295 210 1058 275 1366 215 1111 280 1441 220- 1168 285 1523 225 1229 290 1610 230 1295 295 1703 235 1366 300 1803 2401 1441 305 1909 245 1523 310 2023 250 1610 315 2145 255 1703 320 2274 260- 1803 325 2413 265- 1909 270 2023 275 2145.

280 2274

- 285 2413-Byron Unit 2 Heatup and Cooldown Limit Curves October 1997

23 TABLE 8 Byron Unit 2 Cooldown Data at 12 EFPY Without Margins for instrumentation Errors (Inde:les Vessel Flan 9e requirements of 150*F and 621 psig per 10CFR$0)

Cooldown Curves (3118568588683) 50F 100F Steady State 25F

-T P T P T P T P 60 0 60 0 60 0 60 0 60 505 60 413 60 596 60 551 65 518 65 427 65 606 65 562 70 531 70 442 70 618 70 574 587 75 545 75 459 73 621 75 602 80 563 80 476 80 621 80 85 617 85 576 85 496 85 621 90 621 90 594 90 517 90 621 95 671 95 613 95 539 95 621 100 621 100 621 100 564 100 621 105 621 105 621 105 590 105 621 621 110 621 110 621 110 618 110 115 621 115 621 115 621 115 621 120 621 120 621 120 621 i 120 621 125 621 125 621 125 621 125 621 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 150 973 150 960 150 949 150 942 155 1012 155 1002 155 995 155 998 160 1054 160 1047 160 1045 165 1099 165 1096 170 1147 175 1198 180 1253 185 1313 190 1377 195 1445 200 1518 205 1597 210 1681 215 1772 220 1868 225 1972 230 2083 l 235 2201 240 2328 245 2462 t

October 1997 B; fos: UM 2 Heatup and Cooldown Limit Curves

24 TABLE 9 Ryron Unit 2 Heatup Data at 12 EFPY Without Margins

- for Instrumentation Errors, Using the 1996 App. G Methodology _

(includes Vessel Flange esquirements of 150*F and 621 psq per 10CFR50)

Hestup Curves 0 92210654) ,

100 Heatup Crecat. U mrt Leak Test Lunit T P T P T P

-. 60 0 212 0 191 2000 60 621 212 649' 212 2485 65 621 212- 692 70 621 212 678 75 621 212 669.

80 621 -212' 662 85 621 212 658 90 621 212 '657 95 621 212 659 100 621 212 663 105 621 212 670 110 621 217. -679 115. 621 212 690 120 621 212 703 125 621 212 719 130 621 212 738 135 621 212 758 140 621 212 78i 145 621 212 607 150 621 212 835 150 738 215 866 155 758 220 900 160 781 225 937 165 807 2?#0 977 170 835 '.:35 1021 175 8C6 240 1068 180 900 245 1119 185 937 250- 1174 190 977 255- 1233 195 1021 260 1297 200 1068 265 1366 205 1119 270 1441 210 1174 275 1521 215 1233 280 1607 l

l 220 1297 285 1699 225 1366 290 1798 230 1441 295 1905 235 1521 300 2020 l

! 240 1607 305 2143 l 245 1699 310 2275 l 250 1798 315 2417 255 1905 260 2020 265 2143 270 -2275 275 2417 October 1997 Byron Unit 2 Heatup and Cooldown Limit Curves

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

25 ,

e TABLE 10 Byron Unn 2 Cooldown Data at 12 EFPY Without Margins for instrumentation Errors, Using the 1996 App. G Methodology (includes Vessel Flenge roquaements of 150*F ond 621 Psg per 10CFR50)

CocMown Curves (19dioe64)

SOF 100F Steady State 25F T P T P T P T P 0 60 0 60 0 ,

60 0 60 589 60 538 60 437 60 621 60 601 65 551 65 452 65 621 65 70 565 70 468 70 621 70 614 621 75 561 75 486 75 621 75 621 80 597 80 505 80 621 80 85 614 85 526 ,

85 621 85 621 90 621 90 546  ;

90 621 90 621 621 95 621 95 572 95 6?) 95 621 100 621 100 599 ,

100 621 100 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 621 115 621 115 6?1 115 621 115 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 145 621 145 621 145 621 145 621 621 150 621 150 621 150 621 150 150 1016 150 1006 150 1045 - 150 1028 1074 155 10M6 155 1066 155 1087 155 1123 160 1120 16( 1131 160 1132 160 165 1180- 165 1176 165 1177 170 1232 175 1288 180 1349 185 1413 .

190 1483 195 1556 200 1638 205 1725 210 1818 215 1918 220 2026 225 2141 230 2266 235 2400 October 1997

! Byron Unit 2 Heatup and Cooldown Limit Curves i

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

6 REFERENCES 1 Regulatory Guide 1.99, Revision 2," Radiation Embrittlement of Reactor Vessel Matenals", U.S. Nuclear Regulatory Commission, May,1988, 2 Fracture Toughness Requirements", Branch Technical Position MTEB 6 2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG 0800,1981, 3 WCAP 9517,

• Commonwealth Edison Co. Byron Station Unit i Reactor Vessel Radiation Surveillance Program", J. A. Davidson, July 1979.

4 WCAP-10398, " Commonwealth Edison Co. Byron Station Unit 2 Reactor Vessel 1 Radiation Surveillance Program", L. R. Singer, December 1983.

5 10 CFR Part 50, Appendix G," Fracture Toughness Requirements" Federal Register, '

VrNme 60, No. 243, dated December 19,1995.

6 1989 ASME Boiler and Pressure Vessel (B&PVf Code,Section XI, Appendix G," Fracture Toughness Criteris for Protection Against Failure".

7 1989 Section Ill, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB 2331, " Material for Vessels".

8 ASME Boiler end Pressure Vessel Code,Section XI,

• Rule for Inservice inspection of Nuclear Power Piant Components", Appendix G," Fracture Toughness Criteria for Protection Against Failure", December 1995.

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

P. Lippincott, April 1994.

10 WCAP-14064," Analysis of Capsule W from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program", P. A. Peter, et al., November 1994, 11 WCAP-7924 A, " Basis for lleatup and Cooldown Limt Curves", W. S. Hazetton, et al.,

April 1975.

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

Andrachek, et al., January 1996.

October 1997 Byron Unit 2 Heatup and Cooldown L.imit Curves

27 i

13 Babcock & Wilcox drawing numbers 185265E, Rev. 2;

• Reactor Vessel General Outline".

I 14 WCAP.14824 Rev.1, " Byron Unit i Heatup and Cooldown Limit Curves for Norma!

l Operation and Surveillance Wold Metalintegration for Byron and Braidwood". P. A. i Grendys, Apdf 1997. I 15 WCAP.14063, " Byron Unit 2 Hestup and Cooldown Limit Curves for Normal Operation", i P. A. Peter, November 1994.

i6 l.S Raju and J.C. Newman, Jr., " Stress IntensNy Factor influence Coefficients for Intemal and Extemal Surface Cracks in cylindrical Vessels", in Aspect of Fracture Mechanics in Pressure Vessels and Pipino, ed. S.S. Palusant, and S.G. Sampath, PVP Volume 58, ASME 1982.

+

17. Nucker Design Information Transmhtal, NDIT No. BYR97-346, Rev.0, " Additional data point for weld wire heat number 442002 for incorporation into Table 2 of WCAP 14824 Rev.1", Dated 9/9/97.

d c Byron Unit 2 Heatup and Cooklown Umit Curves October 1997 I

. . . , .