ML20199E853

Heatup & Cooldown Limit Curves for Normal Operation

ML20199E853
Person / Time
Site:	Byron, Braidwood
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-14970, NUDOCS 9711240013
Download: ML20199E853 (33)

Text

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Westinghouse Propriete.ry Class 3

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BRAIDWOOD ESIT 2 HEATCP AND COO:1 DOWS' LIMIT CURVES FOR NORMA:1 OPERATION Westinghouse Energy Systems DO K O 54 p PDR

Westinghouse Propr!rtery Class 3

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HEATCP AND v COOLDOWN LIMIT .,

CU~RVES FOR NORMAL OPERATION

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WESTINGHOUSE NON PROPRIETARY CLASS 3 WCAP 14970 Braidwood Unit 2 Heatup and Cooldown Limit Curves For Normal Operation T. J. Laubham October 1997 Work Performed Under Shop Order COCP 139 Prepared by the Westinghouse Electric Corporation for the Commonwealth Edison Company Approved: - -

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C. H . Boyd, Managdr Equipment & Materials Technology Approved: 4

'D. A . Howell, Mariager Mechanical Systems Integration VESTINGHOUSE ELECTRIC CORPORATION Nuclear Services Division P,0. Box 355 Pittsburgh, Pennsylvania 15230-0355

@ 1997 Westinghouse Electric Corporation All Ri0 hts Reserved

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PREFACE This report has been technivally reviewed and verified by:

E. Terek M.

Bradwood Unit 2 Heatup and Cooldown Limit Curves October 1997

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TABLE OF CONTENTS  :

i L I S T OF F I G U R E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .l t

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- L I S T O F TAB L E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i 1 I NT R OD U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..

2 FRACTURE TOUGHNESS PROPERTIES ............... .. ..... ......... ...................... ... ....... 2 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 F E R E N C E S . .. . . . . . . . . . . . . . . . . . . . . . . - ~ ~ . ~ . . - ~ ~ ~ "- -""" "" - " " " - " " - " " " " " . . . !.

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b Bradwood Unit 2 Hestup and Cooldown Umit Curves October 1997

r T LIST OF FEURES Draidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up 1

' to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for 18 instrumentation Errors). .. .. .

Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates 2

up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for 19 instrumentation Errors).

Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up 3

to 100'F/ht) Applicable for the First 12 EFPY (Without Margins for

. . . . . . . . . . 20 instrumentation Errors; Using 1996 Appendix G Methodology). .. .

4 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 12 EFPY (Without Margins for

. . . . . . . . 21 instrumentation Errors; Using 1996 Appendix G Methodology). ..

Braidwood Unit 2 Heatup and Cooldown 1.imit Curves October 1997

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LIST OF TABLES 1 Calculation of Average Cu and Ni Weight % Values for the Braidwood Unit 2 Base Metals . . . . . . . . . . . . . .. . . . . .. .. .. .. .. . 9 2 Calculation of Average Cu and Ni Weight % Values for the Braidwood Unit 2 Weld Material (Using Braidwood 1 & 2 Chemistry Test Results).. ... .... . . . . . . . . . . . . . 10 3 Braidwood Unit 2 Reactor Vessel Material Properties.. ... .. . . . . . . . . . . . . . . 10 4 Calculation of Chemistry Factors Using Credible Braidwood Units 1 and 2 S urve illa nee C a ps ule Data . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 5 Calculatic.i of Adjusted Reference Temperatures (ART) at 12 EFPY for all Braidwood Unit 2 Reactor Vessel Material (based on credible surveillance capsule data)... 14 6 Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T L o cations f or 12 E F P Y. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 1

Braidwood Unit 2 Heatup Data at 12 EFPY Without Margins for Instrumentation 7

Errors (includes vesset flange requirements of 140*F and 621 psig per 10CFR50). ....... ........... 22 8 Braidwood Unit 2 Cooldown Data at 12 EFPY Without Margins for Instrumentation Errors (includes Vessel flange requirements of 140*F and 621 psig per 10CFR50) . . .. . . . . . . . . . 23 9 Braidwood Unit 2 Heatup Data at 12 EFPY Without Margins for instrumentation Errors. Uaing the 1996 App. G Methodology (includes Vessel flange requirements of 140*F and 621 osig per 10CFR50)...... ... .. . . .. . . . 24 10 Braidwood Unit 2 Cooldown Data at 12 EFPY Without Margins for Instrumentation Errors, Using the 1996 App. G Methodology (includes Vessel flange requirt.ments of 140*F and 621 psig per 10CFR50).. . . . . . . . . . . . . . . . . . . 25 Bradwood Unit 2 Heatup and Cooldown Umit Curves O:tober 1997

's l-1 INTRODUC, TION Heatup and cooldown limit curves are calculated using the adjusted RT,c, (reference nil ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RT,c, of the limiting materialin the core region of the reactor vesselis determined by using the unirradiated reactor vessel matenal fracture toughness properties, estimating the rediatinn induced ART,c,, and adding a margin. The unirradiated RT,c,is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft lb of impact energy and 35 mil lateral expansion (normal to the major working direction) minus 60'F.

RT,e, increases as the material is exposed to f ast-neutron radiation. Therefore, to find the most limiting RT,e, at any time period in the reactor's life, ART,c, due to the radiation exposure cssociated with that time period must be added to the unirradiated RT,e,(IRT,e,). The extent of the shift in RT,e, is enhanced by certain chemical elements (such as copper and nickel) present in reactor ve,ssel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2,

• Radiation Embnttlement of 'teactor Vessel Materials *01. Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRT,e, + ART,c, + margins for uncertainties) at the 1/4T and 3/4T locatiuns, 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.

m Braidwood Unit 2 Heatup and Cooldown Lima Curves October 1997

i 2 FRACTURE TOUGHNESS PROPERTIES The fracture tougMess properties of the ferriti0 materialin 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 Braidwood Unit 2 reactor vessel are presented in Table 3. Credible surveillance data is available for two capsules (Capsules U and X) for Braidwood Unit 2. The post irradiation fracture toughness properties of the reactor vessel surveillance material was obtained directly from the Braidwood Unit 2 Reactor Vessel Radiation Curveillance Program Results"M and was used to calculate chemistry factors (See Table 4). For cil other beltline materials the chemistry factor was calculated per Regulatory Guide 1.99, Revision 2, position 1.1.

Per the requent of the Commcawealth Edison Comoany, the surveillance weld data from the Braidwood Unit 1 and Braidwood Unit 2 surveillance programs"'I has been inteorated !r.

cddition to the credible surveillance weld data from Braidwood Unit 2, credible surveillance weld d:ta is available for two capsules (Capsules U and X) for Braidwood Unit 1. The chemistry fcctor vuues resulting from, the weld mete! integration of the Braidwood Unit 1 and 2 surveillance program results is presented in Table 4 in Section 4 of this report.

Braidwood Untt 2 Heatup and Cooldown Limit Curves October 1997

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/ 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE

! RELATIONSHIPS Appendix G to 10 CFR Part 50,

• Fracture Toughness Requirements *1 specifies tracture toughness requirements for ferritic materials of pressure retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety dunng any condition of normal operation, including anticipated operational occurrences

&;nd system hydrostatic tests, to which the pressure boundary may be subjected over its service lifatime. The ASME Boiler and Pressure Vessel Code forms the basis for these requireinents.

Scction XI, Division 1," Rules for Inserv!ce Inspection of Nuclear Power Plant Components'",

Vessels, contain the conservative methods of analysis.

The ASME approach for calculating the al owable limit curves for various heatup ano cooldawn rates specifies that the total stress intensity factor, K., for the combined thermal and pressure stresses at any time dur'ng heatup or cooldown cannot be greater than the reference stress intensity factor, K,,, for the metal tempe ture at that time. K,, is obtained from the reference frecture toughness curve, defined in Appendix G of the ASME Code,Section XI. The K,, curve is given by the following equation:

""'*'l Ku = 26.78 + 1.23 3

• c'""8 (1) where, K. = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature rte, Therefore, the governing equation for the heatup-ccr'Jown analysis is defined in Appendix G of the ASME Code as follows:

C

• K w+ Kr, < Ku (2) where, K.,, = stress intensity factor caused by membrane (pressure) stress K, = stress intensity factor caused by the W. I gradients K., = function of temperature relative to the RTet of the material C= 2.0 for Level A and Level B service limits C= 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical Bradwood Unit 2 Heatup and Cooldown Limit Curves October 1937

At any time dunng the heatup or cooldown trcnsbnt, the allowabla value of K. is datermined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location. the appropriate value for RTem, 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, K,,, for the reference flaw are computed. From Equation 2, the pressure stress intensity f actors are obtainerd and, from these, the allowable pressures are calcu!sted.

For the calculation of the allowable pressure versua coolant temperature during cooldown, the reference flaw of Appentlix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall becauv the thermal gradients produce tensile stresses at the inside, which increase with increashg cooldown ratt 6. Allowable pressure temperature relations are generated for both steady state anct finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the d limiting pressure is actually dependent on the material temperature at the tip of the assume d flaw. During cooldown, the 1/4T vessellocation is at a higher temperature than the fluid adjacent to the vesselinner diameter. This condition, of course, is not true for the steady state situation. It follows that, at any given reactor coolant tempercture, the AT (temperature) developed during cooldown results in a higher allowable value of K. at the 1/4T location for finite cooldown rates than for steady state operation. Furthermore,if conditions exist so that the increase in allowable value of K. exceeds K., 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 unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure te.nperature relationships are developed for steady state conditions as well cs finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by intemal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the allowable value ci K. for the 1/4T crack during heatup is lower than the allowable value of K for the 1/4T crack during steady state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower allowable K. values do not offset each other, and the pressure temperature curve based Braidwood Unit 2 Heatup and Cooldown L.imit Curves October 1997

F o" 5""dy state conditions no longer repressnts a lower bound of all similar curve

{ rates when the 1/4T flaw is considered Therefore, both cases have to be analyzed in heat s te that at any coolant temperature the lower value of the allowable pressure y

cutated for steady state and finite heatup rates is obtained.

Second poMon of the heatup analysis concerns the calculation of the pressure temperature, .

I' dations for the case in which a 1/4T flaw located at the 14T location from the outsid assumed Uni);e the situation at the vesselinside surface, the thermal gradients establishe t the outside sur ace during heatup produce stresses which are tensile in nature and the t nd to reinforce any pressure stresses present. These thermal stresses are dependent o both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since thermal stresses at the outside are tensile and increase with increasing heatup rates, each bestup rate must be analyzed on an individual basis, ponowing the generation of pressure temperature curves for both the steady state and finite heatup rate situations, the finallimit curves are produced by constructing a composite curve based on a point by point comparisoa of the steady state and finite heatup rate data. At any geven 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 enterion.

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 .,f the closure flange regions must exceed the material unirradiated RT,c, by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is 621 psig for Braidwood Unit 2.

The limiting unirradiated RT,et of 20'F occurs in the vessel flange of the Braidwood Unit 2 reactor vessel, so the minimum allowable temperature of this region is 140'F at pressures greater than 621 psig. This limit is shown in Figures 1 through 4 wherever applicable.

1996 Addenda to ASME Section XI, Appendir. G MethodologyN Appendix G was recently revised to incorporath the most recent elastic solutions for Ki due to pressure and radial thermal gradients. The riew 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 This revision now provides consit tent computational methods for pressure and thermal K, for Braicwood Unit 2 Heatup and Coc' gown Limit Curves October 1997 p

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-r-_--_-__________ _

~ . . . . .

6 thstmal gradier,ts through the vessel wall at any time dunng the trcnsient, Consistent with the

[ onginal version of Appendix G, no contnbution for crack face press'ure is included in the K, due to pressure, and cladding effects are neglected.

Using thesc most recent elastic soliations in the low temperature region will provide some rehef to restrictions associated with reactor operation at relatively low temperatures. Although the relief is relatively smallin terms of absolute allowable pressure, the benefits are substantial because even a small increase in the allowaNe pressure can be a signiftant percentage increase in the operating window at relatively low temperatures. Although implementing this ravision for Braidwood Unit 2 does not result in a change to the steady state curve used as an input to the L. TOP set points, a benefit is achieved at higher cooldown rates in terms of operating window with no reduction in vessel integrity.

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 25 8 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. = 0.8938 for 2s d s 3.464, M. = 3.09 for E > 3.464 and p = intemal pressure, Ri = vesselinner 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. = 0.953x10 x CR x t25, where CR is the cooldown rate in 'F/hr., or for a postulated outside surface defect, K, = 0.753x10'8 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 determined from Fig G 22141. The temperature at any radial distance from the vessel surf ace can be determined from Fig. G 2214 2 for the maximum thermal K, .

Bradwood Unit 2 Heatup and Cooldown Limit Curves October 1997

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)

The maximum thermal Ki relationship and the temperature relationship in Fig G 22141 (a) are applicable only for the conditions given in G 2214.3(a)(1) and (2).

(b) Alternatively, the K, for radial thermal gradient can be calculated for any thermal stress distnbution and at any specified time dunng cooldown for a %-thickness inside surf ace defect using the relationship:

L = (10359Co + 0.6322Ci + 0.4753C: + 0.3855C3)

• 6 (4) or similarly, Kn dunng heatup for a % thickness outside surface defect using the relationship:

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

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

cr(x) = Co + Ci(x / a) + C (x / a)' + C3(x / a)' (6) and x is a varible that represents the radial distance from the appropnate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Braidwood Unit 2 Heatup and Cooldown 1.imit Curves October 1997

CALCULATION OF ADJUSTED REFERENCE TEMPERATU Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each From Ittine region is given by the following expression:

-ees (7)

ART - InitialRTuor + 6 RTuor + Margin

' the reference temperature for the unirradiated m material as defined in paragraph Igal R 5

f ection 111 of the ASME Boiler and Pressure Vessel Code . If meas NS233 for the material in question are not available, generic mean values for that class of n

ay be used if there are sufficient test results to establish a mean and standard gvistion for the class.

is gne mean value of the adjustment h reference temperature caused by irradiation and ARTw should be calculated as follows:

A RTuor = CF

• f* *"' zey To calculate ART,c, at any depth (e g., at 1/4T or 3/4T), the following formula must first be to attenuate the fluence to the specific depth.

f,% = f , ,,* c' * (9) where x inches (vessel,wltline thickness is 8.S' inches"83) is the depth into the vessel wall i

measured from the vesst clad / base metalinterface. The resultant fluence is then placed in I Equation 8 to calculate the ART,et at the specific depth. The calculated surface fluence for Braidwood Unit 2 upper and lower shell forgings and circumferential weld at 12 EFPY is 8.24 x 10" niem'. This fluence value was calculated from '.he surveillance Capsule X analysis prosented in WCAP.14228"M CF ('F)is the chemistry factor, obtained from the tables in Reference 1, using the average

, values of copper and nickel content as calculated 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 chemistry factor for the surveillance weld metal is identical to the best estimate chemistry factor for the vessel weld metal, thus the ratio procedure was not used to adjust the measured l

value of ART,ey.

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All materials in the bettiine region of the Braidwood Unit 2 reactor vessel were considered in determining the limiting material. The calculations to determine the ART values for beltline matenals at 12 EFPY are shown in Table 5. The resulting ART values for all beltline region h

Braidwood Ur42 Heatup and Cooldown Limit Curves October 1997 h

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, maten 15 at the 1/4T and 3/4T locations are summarized in Table 6, where it can ing matenalis the circumferential weld (bpsed on creoible surveillance capsule data).

thel'I 4 and 3/4T ART values for circumferential weld (based on credible integrated W II:nce capsule data) were used in the generation of heatup and cooldown curves able to 12 EFPY.

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• - TABLE 1 Calculation of Average Cu and NiWeight % Values for the Braidwood Unit 2 Base Metals

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Upper Shell Forging Lower Shell Forging 49D9631/49C9041 50D1021/50C97-1 Cu % N% Cu% Ni%

Reference 0.03 0.71 Ref.18 O.057 0.77 Ref. 4 ~

~ 0.49 0.745 Ref.17 ~

0.06 0.75 Ref.19

~

Ref. 20 0.056 0.004 Average 0.03 0.71 0.06 0.77 0.00 0.00 0.046 0.027

~ Standard Deviation .

Note: Averages were originally documented in WCAP 14230f"l, N

4*00d Unit 2 Heatup and Cooldown Limit Curves October 1997 h- -

10 TABLE 2 Calculation of Average Cu and NiWeight Percent Values for the Braidwood Unit 2 Weld Material (Using Braidwood 1 & 2 Chemistry Test Results)'"

Weld Type Cu% Ni%

B&W Weld Qualifications 0.028 0 63 (BAW 2261: WF 501,562) 0 03 0 65 0 04 0.67 Brcewood Untt 1 Weld Date 0.032 0.671 Brcdwood Unit 2 Weld Data *' O.033 0.708 Bes' Estimate Chemistry 0.033 0 666 Standard Deviatiort 0.005 0.029 NOWS:

(a) The we'.d materialin the Braidwood Unit i surveillance program was made of the same wire and Oux as the fettctor vessel girth seam weld. (Weld seam WF 562, Wire Heat No 442011, Flux Type Linde BO, Flux Lot No. 8061)

(b) The weld matenalin the Braidwood Unit 2 surveillance program was made of the same wire and flux as the reactor vessel girth seam weld. (Weld seam WF 562, Wire Heat No. 442011, Flux Type Linde 80, Flur Lot No. 8061)

(c) Repnnted from WCAP 14824 Rev.1'*

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

• RTet (*F)*

Closure Head Flange

• Not Reponed 0.75 -

20'"

Vessel Flange

• 0.07 0.70 -

20'"

Upper Shell Forging 0.03 0.71 20 30 49D963 il49C9041 Lower Shell Forging 0.06 0.77 37 -30 50D102-1/50C971 ,

Circumferential Weld 0.03 0.67 41 40 NOTES:

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

(b) initial RTuo, values are measured values.

(c) Closure head and vessel flange initial RTuot values are used for considenng flange requirements

• for the heatup/cooldown curves.

(d) Ciosure Head Flange nickel value per ref. 21, Vessel Flange values per ref 22 Brc4wtxx! Unit 2 Heatup and Cooldown Limit Curves October 1997

. l.1 TABLE 4 Calculation of Chemistry Factors Using Credible Braidwood Units 1 and 2 Survei: lance Capsule Data Capsule Capsule F P Measured FF* FF2 Material fluence f ARTue, ARTuot Lower Shell Forging U 3.933 x 10" O.741 0 0 0.550 50D1021/50C971 (Tcngentist) 1.067 X 1.126 x 10" 1.033 3 3.099 U 3.933 x 10" O.741 5 <

3.707 0.550 Lower Shell Forging 50D1021/50C971 36.160 1.067 X 1.126 x 10* 1.033 25 (Axist) Sun 42 996 3.234 Chem:stry Factof* = 42.996 + 3.234 = 13.3*F

, Brcidwood i Weld U 3.814 x 10 O.733 10 7.333 0.538 Msta!WF 562*'

X 1.14A x 10 1.038 25 25.95 1.077 Brcidwood 2 Weld U 3 933 x 10" O741 0 0.00 0.550 M;tal WF.562*

W 1.126 x 10" ~ 1.033 20 20.66 1.067 Sum: 53 943 3.232 9

Chemistry Factof* = 53 943 + 3.232 = 16.7'F NOTES' (a) FF = Fluence Factor = f""** *

(b) Braidword Unit 1 ART ie, values were r 5tained from WCAP.14243. The Braidwood Unit 1 capsule fluence values were recalculated using the ENDF/B.V scattenng cross sections in 1994 and are documented in WCAP.14044l.

(c) Braidwood Unit 2 capsule fluence, FF, and 4RT e, values were obtained %m the surveillance Capsule X analysis (WCAP.14228") using the ENDF/B V scattenng cross sections.

(d) Chemistry Factor = I(FF* ART e,) + E(FF8)

Brciowood Unit 2 Heatup and Cooldown Limit Curves October 1997

Expl: nation of Maroin Terms ussd for Brcidwood Unit 2 Wh2n there are "two or more credible surveillance data sets *"I available for Braidwood Unit 2, Rcgulatory Guide 1.99 Rev. 2 (RG1.99R2) Position 2.1 states "To calculate the Margin in this ecse, use Equation 4; the values given there for og may be cut in half". Equation 4 from RG1.99R2 is as follows: M = 2 da ' + o.'

St:ndard Deviation for initial RT,c, Margin Term, e, if the initial RT,c, values are measured values, which t. hey are in the case of Braidwood Unit 2, than o,is equal to O'F, On the other hand, if the initial RT>e, values were not measured, th60 a g:nsric value of 17'F would h*ve been required to be used for o, .

Stendard Deviation for ART,e, Margin Term, ca Per RG1.99R2 Position 1.1, the values of oo are referred to as "28'F for welds and 17'F for b:se metal, except that ca ..ed not exceed 0.50 times the mean value of ART,et." The mean v lue of ART et si defined in R'31.b32 by Equation 2 and defined herein by Equation 8.

Por RG1.99R2 Position 2.1, w .e therr is credible surveillance data, og is taken to be the isssor of % ART,e1 or 14*F (28'F/2)ior welds, or 8.5'F (17'F/2) for base metal. Where ART,c3 agtin is defined herein by Equation 8.

Summary of the Margin Term Sinco o,is taken to be zero when a heat specific measured value of initial RT.et are available (cs they are in this case), the total margin term, based on Equation 4 of RG1.99R2, will be as follows:

e Position 1.1: Lesser of ART,c, or 56'F for Welds Lesser of ART,e, or 34*F for Base Metal e Position 2.1: Lesser of ART,e1 or 28'F for Welds Lesser of ART,e1 or 17'F for Base Metal Brc,dwood Unit 2 Heatup and Cooldown L.imit Curves October 1997

13 The following in a sample calculation of the m:rgin torm for the wald matal et the % T location.

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

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

e From Equation 8 -+ ART,e1 = CF x FF where, CF = 41.0 (R.G. Position 1.1)

= 16.7 (R.G. Position 2.1; i.e. using Surv. Caps. Data)

FF = 0.804 (@ 12 EFPY and Fluence = 8.24 x 10" n/cm')

Therefore, ART,e1 = 32.96 (R.G. Position 1.1)

= 13 43 (R.G. Position 2.1; i.e. using Surv. Caps. Data) e From Equation 4 (of R.G.1.99 R2)-+ Af = 2da' + cr!

where,  % ART,e1 = 16.48 (R.G. Position 1.1)

= 6.72 (R.G. Position 2.1; i.e. using Surv. Caps. Data) e, = 0'F (Initial RT,c1 is Measured) o3 = Lesser of (% ART,et ) or (28'F)

= 16.48 (R.G. Position 1.1) c6 = Lesser of (% ART,er ) or (14'F)

= 6.72 (R.G. Position 2.1; i.e. using Surv. Caps. Data) 2 Therefore, Af = 240' + 16.48 = 32.96 (R.G. Position 1.1)

Af = 240' + 6.72 3

= 13 43 (R.G. Position 2.1; i.e. ushg Surv.

Caps. Data)

Brc6dwood Unit 2 Heatup and Cooldown Limit Curves October 1997

TABLE 5 Calculation of Adjusted Reference Temperatures (ART) at 12 EFPY for a;t Braidwood Unit 2 Reactor Vessel Material (based on credible surveinance capsule data)

Reador Vesset Be%ne Matenal f @ 12 M ART"

%-t (*  %-iFF I ART,,," o, on Region Locafrei ident5caton Cu% M% CF" EFPY (x 10") nti Et FF ,

% T Calculalon 0 495 0 804 -30 16 06 0 8 04 16 06 22 490963-11 0 03 0 71 20 0 0 824 Upper Shes Forgmg 49C904-1 ,

0 804 -35 29.75 0 14 88 29 75 29 5 500102-11 0 06 0 77 37.0 0 824 0 495 Lower Shet Forgog 13 3 0 824 0 495 ~ 0804 -30 10 69 0 545 'i~'55~ 0 -85 Lower shes Forgog

-+ us9S!C Data 40 32.96 0 16 48 32 96 105 9 WF-562 0 03 0 67 41.0 0 824 0 495 0 804 Circ Weld Metal _ _ , , , _ , , , . , , , , _ , _ , _

_LHem442011L 0 6.72 13 43 66 9 16.7 _ 0 824 0 495 0 804 40 13 43 Cire. Weld Metal

-496iSIC Data

% T Calculaton 0.178 0542 -30 10.84 0 5 42 10 84 -8 3 (Jpper Shes Forgog 490953-11 0 03 0 71 20 0 0 824 49C904-1 0.542 -30 20 05 0 10.03 20 05 10 1 Lower She8 Forgog 500102-11 0 06 0.77 37.0 0 824 0 178 50C97-1 0 178 0542 -30 7 21 0 3 61 7 21 -15 6 Lower shes Forgng 13.3 0 824 l

-4 using SIC Data 0.824 0 178 0.542 40 2122 0 11.11 22.22 84 4 WF-562 0 03 0 67 41 0 Circ Weld faetal

&SEL. 16.7 0 824 0.178 0542 40 9 05 . 0 4 53.-

9 05 58 1 Cire. Weld Metal j --, esrg SIC Data ,

NOTES; The Braidwood Urut 2 reactor vessel was thickness is 8.5 inches at the belthne region.

(a) i (b)

ART = 1 + ART,c, + M (This value was tour *d per ASTM E29. Using the *Roundirig Method".)

(c) ART,,, = CF

• FF (d) The CF is integrated between the Braidwood 1 Weld (WF-562, heat # 442011) and the Bradwood 2 Weld (WF-562. Heat # 442011).

October 1997 Braidwood Unit 2 Heatup and Cooldown LiTut Curves

15 TABLE 6 Summary of Adjusted Reference Temperatures (ART) at 1/4T and 3/4T Locations for 12 EFPY Metenal 12 EFPY 1/4T ART 3/4T ART Upper Shell Forging 2.2 -8.3 490963 il49C9041 (RG Position 1(a))

Lower Shell Forging 29.5 10.1 50D1021/50C971 (RG Positson 1(a))

using credible surveillance -B.6 15.6 capsule data (RG Position 2(a))

Circumferential Weld 105.9 84.4 (RG Position 1(a))

using credible surveillance 66,9% 55.1 *'

ccpsule data (RG Position 2(a))

NOTES:

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

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

Braidwood Unit 2 Heatup and Cooldown Umit Curves October 1997

16 5 HEATUP AND COOLDOWN PRESSURE TEMPERATURE LIMIT CURVES Pressure temperature limit curves for normal hestup and cooldown of the pnmary reactor coolcnt system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods"'I discussed in Section 3 and 4 of this report. The 1989 cdition methodology is also presented in WCAP 14040-NP A"", dated January 1996.

Figurcs 1 and 3 present the heatup cu:ves, using ti e 1989 and 1996 Appendix G Methodology respectively, without margins for instrumentation er ors and for a heatup rate of 100'F/hr cpplicable for the first 12 EFPY. Figures 2 and 4 0 esent the cooldown curves, using the 1989 cnd 199G Appendix G Methodology respectively, without margins for instrumentation errors and for cooldown rates up to 100'F/hr applicable for the first 12 EFPY. Allowable combinations of 1:mperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 through 4. This is in addition to other criteria which nut be met before the reactor is made critical. (As a note for Figures 1 through 4, the horizontal axis is the reactor water temperature er moderator temperature and the vertical axis is the calculated cllowable pressure based on through wall stresses.)

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

1.5K > < Ku (10) where, Ko ls the stress intensity factor covered by membrane (pressure) stress, K.= 26.78 + 1.233 e """"* ' "a n, 1

T is the minimum permissible metal temperature, and RTuo, is the metal reference nil-ductility temperature The enticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 5. The prcesure temperature limits or core operation (except for low power physics tests) are that the roactor vessel must be at a temperature equal te or higher than the minimum temperature rcquired for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding pressure temperature curve for heatup and cooldown calculated as described in Section 3 of this report. The minimum temperature for the Brcidwood Unit 2 Heatup and Cooldown Limit Curves October 1997

1I I inscrvice hydrostatic 13ak tasts for the Braidwood Unit 2 reactor vessel et 12 EFPY is 200*F @

2485 psig using the 1989 App. G Methodology and 192'F @ 2485 psig using the 1996 App G Mstnodology. The verticalline drawn from these points on the pressure temperature curve, intersecting a curve 40'F higher than the pretsure temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figurcs 1 through 4 define all of the above limits for ensunng prevention of nonductile f ailure for ths Braidwood Unit 2 reactor vessel. The data points used for the heatup and cooldown pressure temperature limit curves shown in Figures 1 through 4 are presented in Tables 7 through 10.

Add:.ionally, Westinghouse Engineering has reviewed the minimum boltup temperature tcquirements for the Braidwood 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 rc:ctor vessel may be bolted up and pressunzed to 20 percent of the initial hydrostatic test prcssure at the initial rte of the material stressed by the boltup. Therefore, since the mot:t limiting initial RTm value is 20'F (vessel flange), the reactor vessel can be botted up at this 1:mperature. However, based on historical practices and engineenng judgement, W:stinghouse recommends a bolt up temperature of no less than 60'F.

Braidwood Unit 2 Heatup and Cooldown Limit Curves October 1997

18 l

l MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL VELD (using sury. capsule esta)

LIMITING ART VALUES AT 12 EFPY: 1/4T, 66.9'F 3/4T, 58.1'F 2500 ,

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FIEURE 1 B iood Unit 2 Reactor Coolant System Hastup Limitations (Heatup Rates up to 100'F/hr) Ar'plicable for the First 12 EFPY (Without Margins for instrumentation Errors)

Braidwood Unit 2 Heaton and Cooldown Limrt Curves October 1997

____ _ _ _ _ _ _ _ _ _ J

19 l

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FIGURE 2 , Brs;dwood Unit 2 Reactor Coo' ant Systern Cooldown Limitations (Cooldown Rates up to 100*F/hr) /.pplicable for the First 12 EFPY (Without Margins for instrumentation Errors)

Braidwood Unit 2 Heatup and Cooldown Limit Curves October 1997

=

20 MATERIAL PROPERTY BASIS -

LIMITING MATERIAL: CIRCUMFERENTIAL WELD (using surv. c.apsule data) ,

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L 1

FIGURE 3' Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100*F/hr) Applicable for the First 12 EFPY (Without Margins for l Instrumentation Errors, Using 1996 Appendix G Methodology)

Braidwood Unit 2 Heatup and Cooldown Limit Curves October 1997

21

' MATERIAL PROFERTY BASISL ,

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

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L FIGURE 4 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown l

Rates up to 100'F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology)

Braidwood Unit 2 Heatup and Cooldown Limit Curves October 1997

22

(

TABLE 7 Braidwood Unit 2 Heatup Data at 12 EFPY Without Margins for Instrumentation Errors (includes Vessel Flange requirements of 140*F and 621 psig per 10CFR50)

Hestup Curves (Con 6guration # 1318789467) 100 Heatap Cnticality Limrt Leak Tect Limit T P T P T P -

200 0 179 2000 60 0 200 644 200 2485 60 621 65 621 200 644 85 621- 200 644 90 621 .200 644 95 621 200 644 100 621 200 644 105 621 200 545 110 621 200 649 115 621 200 656 120 621 200 665 125 621 200 676 130 621 200 689 135 621 200 705 140 621 200 724 140 705 200 745 145 724 200 768-150 745- 200 794 155 768 .

205 823 160 794 210 854 165 823 215 888 170 854 220 926 175 888 225 966 180 926 230 1010 185 966 235 1058 190 1010 240 1109 195 1058 245 1164 200 1109 250 1224 205 1164 255 1288 210 1224 260 1357 215 1288 ?65 1432 220 1357 270 1511 225 1432 275 1596 230- 1511 280 1688 235 1596 285 1786 240 1688 290 1891 245- 1786 295 2002 l 250 1891 300 2122

255 2002 305 2250 260 2122 310 2386 265 2250 270 2386 Braidwood Unit 2 Heatup and Cooldown Limr " October 199f

_. _ _ . ~ . . _

TABLE 8-Braidwood Unit 2 Cooldown Data at 12 EFPY Without Margins for instrumentation Errors (includes Vessel Flange requirements of 140*F and 621 psig per 10CFR;0)

Cooldown Curves (Configuration # 1318789467)

Steady State 25F_ 50F 100F T P T P T P T P 60 0 60 0 60 0. 60 0 60 621 60 605 60 564 60 482 65 621 65 620 65 580 65 502 70 621 70 S21 70 598 70 523-75 621 75 621 75 617 -75 545 80 621 80 621 80 621 80 570 85 1 621 85 621 85 621 85 596 '

-90 621 90 621 90 621 90 621 95 C21 95 621 95 621 95 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 110: 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 140 621_ 140 621 140 621 140 1060 140 1054 140 1053 145 1105 145 1103 150 1154 155 1206 160 1262 165 1322 170 1386 175 1455 180 1529 185- 1609 190 1694 195 1785 200 1882 205 1987

.210 2099 215 2218 220 2346 225 2482 l

l l

! Braidwood Unit 2 Heatup and Cooldown Limit Curves October 1997

TABLE 9

- Braidwood Unit 2 Heatup Data at 12 EFPY Without Margins for instrumentation Errors, Using the 1996 App. G Methodology (includes Vessel Flange requirements of 140*F ud 621 psig per 10CFR50)

Heatup Curves (Computer Run # 107854252) 100 Heatup - Cnticality Limit Leak Test 8.imit T P T- P T P 192 0 170 2000 60 0 192- 706 192 2485 60 621 65 621 192' 741 85 621 192 729 90 621 192 720 95 621 192 715 100 621 192 713 105 621 192 714 110 621 192 719 115- 621 192 726 120- 621 192 735 125 621 192 748 130- 621 192 763 135 621- 192 -780

-140 621 192 801 140 780 192 824 145 801 195 850 150 824 200- 879 155 850 205 911 160 879 210 946 165 911 215 984 .

170 946 220 1026 175 984 225 1071 180 .1026 230 1120 185 1071 235 1174 190 1120 240 1231 195 1174 245 1294 200 1231 250 1361 205 -1294 255 1433 210 1361 260 1511 215 1433 265 1595 220 1511 270 1686 l

225 1595 275 1783 230' 1686 280 1838 235 1785 285 2000 240 1888 290 2121 245 2000 295 2251

-250 2121 300 2391 l 255 2251 l 260 2391 l~

Braidwood Unit 1 Heatup and Cooldown Limrt Curves October 1997

TABLE 10 Braidwood Unit 2 Cooldown Data at 12 EFPY Without Margins for instrumentation ' ' ors, Using the 1996 App. G Methodolog)

(includes vessel Flary wrerrsnts of 140'F and 621 psig per 10CFR$0)

Cooldown Curves (Computer Run # 107854252) 25F SOF 100F

- Steady State T P T P T P T P W 0 O- 0  % 0 M 0 621 60 601 60 511 60 621 60 621 65 619 65 532 65 621 65 621 70 621' 70 555 70 621 70 621 75 621 75 579 75 621 75 621 80 621 80 605 80 621 80 621 85 621 85 621 85 621 85 90 621 90 621 90 621 90 621~

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 G21 115 621 115 621 115 621 11b 621 120 621 120 621 120 621 120 125 621 125 621 125 621 125 621 621 130 621 130 621~

130 621 130 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 1138 140 1131 140 1128 145 1187 145 1184 145 1186 150 1240 155 1297 160 1357 165 1423 170 1493 175 1569 180- 1650 185 1738 190 1832 195 1933 200 2041 205 2158 210 2284 215 2419 l

l Braidwood Unit 2 Heatup and Cooldown Limit Curves October 1997

6- REFERENCES - . ..

1 - Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Re' actor Vessel

~

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

-2 Fracture Toughness Requirements", Branch Technical Position MTEB 5-2, Chapter 5.3.2 -

in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power

- Plants, LWR Edition, NUREG-0800,1981, -

, 3 WCAP 9807, " Commonwealth Edison Co. Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program", S.E. Yanicoko, et. al., February 1981.

4 WCAP 11188, " Commonwealth Edison Co. Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program"i L. R. Singer December 1986, e

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

' . Volume 60, No. 243, dated December 19,1995.

6 1989 ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix G. " Fracture

- Toughness Criteria for Protection Against Failure",

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

8' ASME Boiler and Pressure Vessel Code,Section XI, " Rule for Inservice inspection of Nuclear Power Plant Components", Appendix G, " Fracture Tcughness Cnteria for Pro *ection Against Failure", December 1996.

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

P. Lippincott, April 1994.

10 WCAP-14228, " Analysis of Capsule X from the Commonwealth Edison Company 4 Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program", P. A. Peter, March .

1995.

11 WCAP 7924-A, " Basis for Heatup and Coofdown Limit Curves", W. S. Hazelton, et al.,

' April 1975.

12 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.

1'3 Babcock & Wilcox drawing numbers 185328E Rev. 2; " Longitudinal Sections" Breedwood Unit 2 Heatup and Coo:Jown Limit Curves Ociober 1997

['

w - ,- , , - - + , , , . . w. ;-._ e . , - - - --w . . . - , .- - , _ # , - ,-.-,,,_,._-,m, ,, , . -m_. .- e

14 WCAP 14824 Rev.1, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metalintegration for Byron and Braidwood", P. A.

Grendys, April 1997, 15 WCAP 14230, "Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation", P. A. Peter, March 1995, 16 l.S Raju and J.C. Newman, Jr., " Stress Intensity Factor influence Coefficients for intemal and External Sittface Cracks in Cylindrical Vessels", in Aspect of Fracture Mechanics in Pressure Vessels and Piping, ed. S.S. Palusamy and S.G. Sampath, PVP-Volume 58, ASME 1982, 17 WCAP 12845," Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program", E. Terek, et. al.,

March 1991.

18. Material Test Results, Upper Shell, The Japan Steel Works, Ltd., JSW Job No. FN3-4274, IR No. 4274 309(1), December 13,1974.

19. Material Test Results, Lower Shell, The Japan Steel Works, Ltd., JSW Job No. FN3-4274, IR No. 4274-410(1), May 30,1975.

20.- Analytical Request #15482, "Braidwood Nuclear Plant, Unit 2, Irradiated Low Alloy Steel Reactor Surveillance Dosimetry", L. Kardos, October 27,1994.

21. United States Steel Corp. Material Test Report for Heat Number 2031-V-1, Dated May 17,1976.

22. Bethlehem Steel Corp. Material Test Report Number 654 for Heat Number 124P455, Dated June 27,1975.

l l

l BraKlwood Unit 2 Heatup and Cooldown Limit Curves October 1997