ML20153G277
| ML20153G277 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 09/30/1998 |
| From: | Christopher Boyd, Laubham T, Trombola D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20153G260 | List: |
| References | |
| WCAP-15079, WCAP-15079-R01, WCAP-15079-R1, NUDOCS 9809300022 | |
| Download: ML20153G277 (38) | |
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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15079, Revision 1 Wolf Creek Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubham September 1998 Work Performed Under Shop Order K6TP-139 Approved: 6 C. H. Boyd, Manager Equipment & Materiais Technology Approved: D. M. Trombola, Manager Mechanical Systems Integration Westinghouse Electric Company Energy Systems P.O. Box 355 Pittsburgh, PA 15230-0355 C1998 Westinghouse Electric Company All Rights Reserved
1 iii 9 TABLE OF CONTENTS LIST OF TABLES. . .. . . . . . . .. . .. .. . iv l 1 LIST OF FIGURES. . . . . . . . . . . . .. vi l PREFACE . . . . .. . . . . . . . . . . . . . . . . . . . . . . vii ; EXECUTIVE
SUMMARY
(OR) ABSTRACT.. .. . .. . . .. . .. viii l 1 1 1 1 INTRODUCTION .. . . . .... . . . . .. . . . . . . . . . . .. . . 1-1 I 2 PURPOSE . . . . . . . .. . . . . . . . . . . . . . . . . 2- 1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS.. . .3-1 1 I l
'4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE . . . . . . . . 4-1 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE CURVES . . ... .. 5-1 6 REFERENCES . . . . . .. . . . . . . . . 6-1 Revision 1
t iv LIST OFTABLES Table 4-1 Summary of the Peak Pressure Vessel Neutron Fluence Values at 20 EFPY used for the calculation ofART Values (n/cm 2, E > 1.0 MeV)... . . . . .. . . . . . . . 4-2 Table 4-2 Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the calculation ofART Values (n/cm2 , E > 1.0 MeV).. . .. . . . .4-3 l Table 4-3 Calculated Integrated Neutron Exposure of the Wolf Creek Surveillance Capsules Tested to Date. . . . . . . . . . . . . . . . . .. . . . . . . . 4-4 Table 4-4 Measured 30 ft-lb Transition temperature Shifts of the Beltime Materials Contained in the Surveillance Program ... .. . . . . . . ..4-5 Table 4-5 Reactor Vessel Beltline Material Unirradiated Toughness Properties. . . .4-6 Table 4-6 Calculation of Chemistry Factors using Wolf Creek Sun eillance ! Capsule Data. . . . . . . . . . . .. . . . . . . . . 4-7 i Table 4-7 Summary of the Wolf Creek Reactor Vessel Beltline Material Chemistry Factors Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1. . . 4-8 Table 4-8 Summary of the Calculated Fluence Factors used for the Geeneration of the 20 EFPYHeatup and Cooldown Curves. . . . . . . . . . . .. .. . . . .4-9 Table 4-9 Summary of the Calculated Fluence Factors used for the Geeneration of the 32 EFPY Heatup and Cooldown Cun'es . . . . . ... . . .. . . .. 4-10 Table 4-10 Calculation of the ART Values for the 1/4T Location @ 20 EFPY.. . . .4-11 1 Table 4-11 Calculation of the ART Values for the 3/4T Location @ 20 EFPY.. .. . .4-12 Table 4-12 Calculation of the ART Values for the 1/4T Location @ 32 EFPY.. . . . . .... .4-13 Table 413 Calculation of the ART Values for the 3/4T Location @ 32 EFPY.. . ...( 14
. Table 414 Summary of the Limiting ART Values used in the Generation of the Wolf Creek Unit 1 Heatup/Cooldown Curves.. . . . .4-15 i
Table 5-1 Wolf Creek Heatup Data at 20 EFPY Without Margins for Instrumentation Errors . . . . . . . . . . . . .. .5-5 l Table 5-2 Wolf Creek Cooldown Data at 20 EFPY Without Margins for i Instrumentation Errors . . . .. . . . .. . . ..5-7 r Revision 1
v LIST OF TABLES - (Continued) Table 5-3 Wolf Creek Heatup Data at 32 EFPY Without Margins for Instrumentation Errors . . . . . . . . . . . . .. . . . . . . . . . . . . . . . ..,5-11 Table 5-4 Wolf Creek Cooldown Data at 32 EFPYWithout Margins for Instrumentation Errors . . ......... . . . . . . . . . . . . . . . . . . . . . .. .5-13 9 P b 4 l I I I l Revision 1
vi LIST OF FIGURES Figure 5-1 Wolf Creek Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100*F/hr) Applicable to 20 EFPY (Ethout Margins for Instmmentation Error) . . .5-3 Figure 5-2 Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,20,40,60 and 100*F/hr) Applicable to 20 EFPY (Without Margins for Instrumentation Error) . . . . . . . . . .5-4 Figure 5-3 Wolf Creek Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100 F/hr) Applicable to 32 EFPY (Without Margms for Instrumentation Error) . .5-9 Figure 5-4 Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,20,40,60 and 100*F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Error) .. . . . . .. . .5-10 8 4 Revision 1
vu PREFACE This report has been technically reviewed and verified by: r Reviewer: Ed Terek M O Revision 1
viii EXECUTIVE
SUMMARY
The purpose of this report is to generate pressure-temperature limit curves for Wolf Creek for normal operation at 20 and 32 EFPY using the methodology from the 1989 ASME Boiler and Pressure Vessel Code, Section XI, Appendix G Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T location. The 1/4T and 3/4T values are summarized in Table 4-14 and were calculated using the lower shell plate R2508-3 (i.e. The limiting beltline region material). The pressure-temperature limit curves were generated for heatup rates of 60 and 100'F/hr and cooldown rates of 0,20,40,60 and 100 F/hr. These curves can be found in Figures 5-1 through 5-4. i I i \ Revision 1
l 11 1 INTRODUCTION
, Heatup and cooldown limit curves are calculated using the adjusted RTmr (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTmr of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTer, and adding a margin. The unirradiated RTmr is designated as the higher ofeither the drop weight l nil-ductility transition temperature (NDTf) or the temperature at which the material exhibits at least 50 ft-lb ofimpact energy and 35-mil lateral expansion (normal to the major working direction) minus 60 F.
RTmyr increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTer t aany time period in the reactor's life, ARTmr due to the radiation exposure associated with that time period must be added to the unirradiated RTer(IRTer). The extent of the shift in RTer is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials"N Regulatory Guide 1.99, Revision 2, is used for the calculation ofAdjusted Reference Temperature (ART) values (IRTer + ARTer + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of I i 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 for normal operation. l l l' .e. e I 4 Introduction Revision 1 l . 1
2-1 2 PURPOSE
, The Wolf Creek Nuclear Operating Corporation has contracted Westinghouse to analyze surveillance capsule V from the Wolf Creek reactor vessel, As a part of this analysis Westinghouse generated new heatup and cooldown curves for 20 and 32 EFPY. The heatup and cooldown curves were generated without margins for instrumentation errors. The curves include a hydrostatic leak test limit curve from 2485 to 2000 psig and pressure-temperature limits for the vessel flange regions per the requirements of 10 CFR Part 50, Appendix G t2),.
The purpose of this report is to present the calculations and the development of the Wolf Creek heatup and cooldown curves for 20 and 32 EFPY. This report documents the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2DI, for all the beltline materials and the development of the heatup and cooldown pressure-temperature limit curves for normal operation. l i l I I 1 4 i Purpose Revision 1
3-1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements"[2] specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary oflight water nuclear power reactors to provide adequate n'argins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements. Section XI, Division 1, " Rules for Insenice Inspection of I Nuclear Power Plant Components", Appendix G '1, contains the conservative methods of analysis. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K ,i for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Ku, for the metal temperature at that time. Ku is obtained from the reference fracture toughness curve, defined in Appendix G of the ASME Code, Section XI. The Ku curve is given by the following equation: Ku = 26.78 + 1.233
- eto ouxr-arenoa (3) where, Ku = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTer Therefore, the goveming equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C
- K1m+ Kn < Ku (2) where, Km=
i stress intensity factor caused by membrane (pressure) stress Ka = stress intensity factor caused by the thermal gradients Ku = function of temperature relative to the RTun of the material C= 2.0 for Level A and Level B senice limits C= 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical Criteria For Allowable Pressure-Temperature Relationships Revision 1
3-2 At any time during the heatup or cooldown transient, Ku is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RTer, and the reference fracture toughness curve. The th nal stresses resulting from the temperature gradients through the vessel , wall are calculated and then the corresponding (thermal) stress intensity factors, Ka, for the reference flaw
, are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw ofAppendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate ofinterest. He use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of Ku at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in Ku exceeds Ka, the calculated allowable pressure during cooldown will be greater than the steady-state value. The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall, ne heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the Ku for the 1/4T crack during heatup is lower than the Ku 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 Kuvalues do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. . Herefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value
' of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. 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 Criteria For Allowable Pressure-Temperature Relationships Revision 1
l 3-3 present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with ! increasing heatup rates, each heatup rate must be analyzed on an indisidual basis. 1 Following the generation ofpressure-temperature curves for both the steady state and fmite heatup rate ; situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion. 1 10 CFR Part 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange l regions. This rule states that the metal temperature of the closure flange regions must exceed the material unitradiated RTm by at least 120 F for normal operation when the pressure exceeds 20 percent of the presenice hydrostatic test pressure (3106 psig), which is 621 psig for the Wolf Creek reactor vessel. The limiting unirradiated RTmr of 20 F (Table 4-6) occurs in the closure head and vessel flange of the i Wolf Creek reactor vessel, so the muumum allowable temperature of this region is 140 F at pressures greater than 621 psig with no margin for uncertainties. l l l l l l l i l l Criteria For Allowable Pressure-Temperature Relationsidps Revision 1
4-1 ,
- 4. CALCULATION OF ADJUSTED REFERENCE TEMPERATURE
,, From Regulatory Guide 1.99, Raision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression: .
ART = InitialRTmr + ARTmr + Margin (3) Initial RTer is the reference temperature for the unirradiated material as defmed in paragraph NB-2331 of . MI
- Section III of the ASME Boiler and Pressure Vessel Code . Ifmeasured values ofinitial RTwr for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.
ARTer is the mean value of the adjustment in reference temperature caused by irradiation and should be l calculated as follows: 1 A RTer = CF
- f("#"" '# (4)
I To calculate ARTer taany depth (e.g., at 1/4T or 3/4T), the following formula must first be used to j attenuate the fluence at the specific depth. j
/u,,% = / 9,,
- e6 "* (5) where x inches (vessel beltline thickness is 8.63 inchesDI) is the depth into the vessel wall measured from 4
the vessel clad / base metal interface. The resultant fluence is then placed in Equation 4 to calculate the ARTer t athe specific depth. The Westinghouse Radiation Engineering and Analysis group evaluated the vessel fluence projectionsl71 and the results are presented in Section 6 of WCAP-15078 . The evaluation used the ENDF/B-VI scattering cross-section data set. This is consistent with the methods presented in WCAP-14040-NP-A, ;
" Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and !
Cooldown Limit Curves"DI. Tables 4-1 r.nd 4-2, herein, contain the best estimate vessel surface fluence values along with the Regulatory Guide 1.99, Revision 2,1/4T and 3/4T calculated fluences used to calculate the ART values for all beltline materials in the Wolf Creek reactor vessel. Additionally, the measured surveillance capsule fluence values are presented in Table 4-3. , i I i 4 Calculation of Adjusted Reference Temperature Revision 1
4-2 TABLE 4-1 - j Summary of the Peak Pressure Vessel Neutron Fluence Values at 20 EFPY used for the Calculation ofART Values (n/cm2 , E > 1.0 MeV) Material Surface %T %T Intermediate Shell Plate R2005-1 1.265 x 10 7.54 x 10" 2.68 x 10" Intermediate Shell Plate R2005-2 1.265 x 10 7.54 x 10" 2.68 x 10" Intennediate Shell Plate R2005-3 1.265 x 10 7.54 x 10" 2.68 x 10" Lower Shell Plate R2508-1 1.265 x 10 7.54 x 10" 2.68 x 10" Lower Shell Plate R2508-2 1.265 x 10" 7.54 x 10" 2 68 x 10" Lower Shell Plate R2508-3 1.265 x 10 7.54 x 10" 2.68 x 10" Intermediate and Lower Shell Longitudinal 6.82 x 10" 4.06 x 10" 1.44 x 10" Weld Seam 101-124A & 101-142A (90' Azimuth) Intermediate and lower Shell Longitudinal 1.265 x 10" 7.54 x 10" 2.68 x 10" Weld Seam 101-124B,C and 101-142B,C (210 &330 Azimuth) Intermediate to Lower Shell 1.265 x 10" 7.54 x 10" 2.68 x 10" Cirumferential Weld Seam 101-171 P Calculation of Adjusted Reference Temperature Revision 1
l 4-3 l I l i TABLE 4-2 Summary of the Peak Pressure Vessel Neutron Fluence Values l 2
,, at 32 EFPY used for the Calculation of ART Values (n/cm , E > 1.0 MeV)
Material Surface %T %T ! Intermediate Shell Plate R2005-1 2.00 x 10" 1.19 x 10" 4.23 x 10" Intermediate Shell Plate R2005-2 2.00 x 10" 1.19 x 10" 4.23 x 10" f i Intermediate Shell Plate R2005-3 2.00 x 10" 1.19 x 10" 4.23 x 10" Lower Shell Plate R2508-1 2.00 x 10" 1.19 x 10" 4.23 x 10" Lower Shell Plate R2508-2 2.00 x 10" 1.19 x 10" 4.23 x 10" i 1 Lower Shell Plate R2508-3 2.00 x 10" 1.19 x 10" 4.23 x 10" l Intermediate and Lower Shell Longitudinal 1.06 x 10" 6.32 x 10" 2.24 x 10" Weld Seam 101-124A & 101-142A (90* Azimuth) Intermediate and Lower Shell Longitudinal 2.00 x 10" 1.19 x 10" 4.23 x 10" Weld Seam 101 124B,C and 101-142B,C (210 & 330 Anmuth) Intermediate to Lower Shell 2.00 x 10" 1.19 x 10" 4.23 x 10" Cirumferential Weld Seam 101-171 l i l l 1 h Calculation of Adjusted Reference Temperature Revision 1
4-4 TABLE 4-3 Calculated Integrated Neutron Exposure of the Wolf Creek .; Surveillance Capsules Tested to Date . Capsule Fluence U 3.429 x 10" n/cm2 , (E > 1.0 MeV) Y 1.308 x 10" n/cm2 , (E > 1.0 MeV) V 2.528 x 10" n/cm2 , (E > 1.0 MeV) 1 l 2 Margin is calculated as, M = 2Ja ,g2 The standard deviation for the initial RTermargin term, o,, is 0 F when the initial RTmr is a measured value, and 17 F when a generic value is used. The standard j deviation for the ARTermargm term, c6, is 17 F for plates when surveillance capsule data is not used and l 8.5'F for plates when surveillance capsule data is used. For welds, ei is 28 F when surveillance capsule j data is not used and 14'F when surveillance capsule data is used. In addition, c3 need not exceed one-half l the mean value of ARTer. i i i I Calculation of Adjusted Reference Temperature Revision 1 l
4-5 Contained in Table 4-4 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline M materials , These measured shift values were obtained using CVGRAPH, Version 4.l l'l, which is a hyperbolic tangent cun'e-fitting program. TABLE 4-4 Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program Material Capsule Measured 30 ft-lb Transition Temperature Shift" Lower Shell Plate R2508-3 U 36.46 F Y 16.03 F (Longitudinal Orientation) V 52.03 F 1 Lower Shell Plate R2508-3 U 23.79 F Y 35.39 F (Transverse Orientation) V 54.53'F Suneillan'e Program U 27.21 F Weld Metal Y 45.09 F V 46.33 F U 58.41 F Heat Affected Zone Y 12.98*F V 55.91 F Notes. (a) Calculated using measured Charpy data and plotted using CVGRAPH ! 'l !O l 1 l l Caledation of Adjusted Reference Temperature Revision 1 1
~
4-6 Table 4-5 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RTmr fothe beltline materials and vessel flanges. The weight percent values of Cu and Ni given in Table 4-5 were used to generate the calculated chemistry factor (CF) values based on Tables 1 and 2 of . Regulatory Guide 1.99, Revision 2, and presented in Table 4-7. Table 4-6 provides the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Position 2.1, which are also summarized in Table 4-7. TABLE 4-5 Reactor Vessel Beltline Material Unirradiated Toughness Propertiestml Material Description Cu(%) Ni(%) Initial RTmr* Closure Head Flange R2504-1 -- 0.66 20 F Vessel Flange R2501-1 -- 0.70 20 F Intermediate Shell Plate R2005-1 0.04 0.66 -20 F Intermediate Shell Plate R2005-2 0.04 0.64 -20 F . Intermediate Shell Plate R2005-3 0.05 0.63 -20 F Lower Shell Plate R2508-1 0.09 0.67 0F Lower Shell Plate R2508-2 0.06 0.64 10 F Lower Shell Plate R2508-3 0.09 0.58 40 F Intermediate and Lower Shell 0.04 0.08 -50 F LongitudinalWeld Seams N Intermediate to Lower Shell 0.04 0.08 -50 F CircumferentialWeld ScamN Surveillance Program Weld Metal N 0.07 0.10 --- Notes: (a) The initial RTervalues for the plates and welds are based on measured data. (b) All welds, including the surveillance weld, were fabricated with weld wire heat number 90146. The , intermediate to lower shell circumferential weld seam 101-171 was fabricated with Flux Type 124 Lot Number 1061. The intermediate shcIl longitudinal weld seams 101-124A, B & C and lower shell . l longitudinal weld seams 101-142A, B, & C were fabricated with Flux Type 0091 Lot 0842. The surveillance weld metal was fabricated with weld wire heat number 90146, Flux Type 124 Lot number , 1061. Per Regulatory Guide 1.99, Revision 2, " weight percent copper " and " weight percent nickel" are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or fcr weld samples made with the weld wire heat number that matches the critical vessel weld." The surveillance weld metal was made with the same weld wire heat as all of the vessel beltline weld seam and is therefore representative of all of the beltline weld seams. Calculation of Adjusted Reference Temperature Revision 1
4-7 t l TABLE 4-6 Calculation of Chemistry Factors using Wolf Creek Surveillance Capsule Data
. Material Capsule Capsule f* FF*) ARTer(*) 2 FF*ARTor FF {
! l Intermediate Sbil U 0.3429 0.705 36.46 F 25.70 F 0.50 Plate R2508-3 Y 1.308 1.075 16.03 F 17.23 F 1.16 l
(Longitudinal) V 2.528 1.249 52.03*F 64.99 F 1.56 Intermediate Shell U 0.3429 0.705 23.79 F 16.77 F 0.50 Plate R2508-3 Y 1.308 1.075 35.39 F 38.04 F 1.16 (Transverse) V 2.528 1.249 54.53*F 68.11 F 1.56 l SUM 230.84 F 6.44 2 CFn25cs.3 = Z(FF
- RTmr) + Z(FF ) = (230.8) + (6.44) = 35.8 F Beltline Region U 0.3429 0.705 19.75 F(d) 13.92 F 0.50 Weld Metal Y 1.308 1.075 32.74 F(d) 35.20 F 1.16 l
V 2.528 1.249 33.64 F") 42.02*F 1.56 SUM 91.14 F 3.22 2 CFw.ia = I(FF
- RTmr) + Z(FF ) = (91.14) + (3.22) = 28.3 F Notes:
(a) f = Measured fluence from capsule V dosimetry analysis results G , (x 10" n/cm2, E > 1.0 MeV). (b) FF = fluence factor =C f ' " 80 (c) ARTervalues are the measured 30 ft-lb shift values. l (d) The surveillance weld metal ARTer values have been adjusted by a ratio factor of 0.726. l L l o l l Calculation of Adjusted Reference Temperature Revision 1 l l l
4-8 TABLE 4-7 Summary of the Wolf Creek Reactor Vessel Beltline Material Chemistry Factors . Based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Material Chemistry Factor Position 1.1(*) Position 2.1(') Intermediate Shell Plate R2005-1 26.0*F - Intermediate Shell Plate R2005-2 26.0 F - Intermediate Shell Plate R2005-3 31.0 F - Lower Shell Plate R2508-1 58.0 F - Lower Shell Plate R2508-2 37.0 F - Lower Shell Plate R2508-3 58.0 F 35.8 F-Intermediate and Lcwer Shell Longitudinal 31.6 F 28.3 F Weld Seam 101-124A & 101-142A*) Intermediate and Lower Shell Longitudinal 31.6 F 28.3 F Weld Seam 101-124B,C and 101-142B,C*) Intermediate to Lower Shell Cirumferential 31.6*F 28.3'F Weld Seam 101-171*) Surveillance Program Weld Metal *) 43.5 F - Notes: (a) Regulatory Guide 1.99, Revision 2, Position 1.1 or Position 2.1 methodology. (b) All welds, including the surveihance weld, were fabricated with weld wire heat number 90146. The intermediate to lower shell circumferential weld seam 101-171 was fabricated with Flux Type 124 Lot Number 1061. The intermediate shell longitudinal weld seams 101-124A, B & C and lower shell longitudinal weld seams 101-142A, B, & C were fabricated with Flux Type 0091 Lot 0842. The sutveillance weld metal was fabricated with weld wire heat number 90146, Flux Type 124 Lot number 1061 , Calculation of Adjusted Reference Temperature Revision 1
4-9 Contained in Tables 4-8 and 4-9 are sununaries of the fluence factors (FF) used in the calculation of adjusted reference temperatures for the Wolf Creek reactor vessel beltline materials for 20 EFPY and 32 EFPY. TABLE 4-8 Summary of the Calculated Fluence Factors used for the Generation of the 20 EFPY Heatup and Cooldown Curves Material %Tf % T FF') %Tf % T FF*)
-l (n/cm', E > 1.0 (n/cm 2, E >1.0 MeV) MeV)
Interrnediate Shell Plate R2005-1 7.54 x 10" 0.92 2.68 x 10" 0.64 Intermediate Shell Plate R2005 2 7.54 x 10" 0.92 2.68 x 10" 0.64 Intermediate Shell Plate R2005-3 7.54 x 10" 0.92 2.68 x 10" 0.64 Lower Shell Plate R2508-1 7.54 x 10" 0.92 2.68 x 10" 0.64 Lower Shell Plate R2508-2 7.54 x 10" 0.92 2.68 x 10" 0.64 Lower Shell Plate R2508-3 7.54 x 10" 0.92 2.68 x 10" 0.64
)
Intermediate and Lower Shell 4.06 x 10" 0.75 1.44 x 10" 0.49 i LongitudinalWeld Seam j 101-124A & 101-142A (90* Azimutl^ Intermediate and Lower Shell 7.54 x 10" 0.92 2.68 x 10" 0.64 LongitudinalWeld Seam 101-124B,C and 101-142B,C (210 & 330* Azimuth) Intermediate to Lower Shell 7.54 x 10" 0.92 2.68 x 10" 0.64 CirumferentialWeld Seam 101-171 Notes: (a) Fluence Factor at the 1/4T vessel thickness location. l (b) Fluence Factor at the 3/4T vessel thickness location. l i i, 8 Calculation of Adjusted Reference Temperature Revision 1
4-10 TABLE 4-9 Summary of the Calculated Fluence Factors used for the Generation of the . 32 EFPY Heatup and Cooldomi Curves Material %Tf %TFF" %Tf % T FF" ' 2 (n/cs', E > 1.0 (n/cm , E > 1.0 MeV) MeV) Intermediate Shell Plate R2005-1 1,19 x 10 19 1.05 4.23 x 10" 0.76 Intermediate Shell Plate R2005-2 1.19 x 10 1.05 4.23 x 10" 0.76 Intermediate Shell Plate R2005-3 1,19 x 10 19 1.05 4.23 x 10" 0.76 Lower Shell Plate R2508-1 1.19 x 10" 1.05 4.23 x 10" 0.76 Lower Shell Plate R2508-2 1.19 x 10" 1.05 4.23 x 10" 0.76 Lower Shell Plate R2508-3 1.19 x 10" 1.05 4.23 x 10" 0.76 Intermediate and Lower Shell 6.32 x 10" 0.87 2.24 x 10" 0.60 LongitudinalWeld Seam 101-124A & 101-142A(90 Azunuth) Intermediate and Lower Shell 1.19 x 10" 1.05 4.23 x 10" 0.76 LongitudinalWeld Seam 101-124B,C and 101-142B,C (210* & 330' Azimuth) Intermediate to Lower Shell 1.19 x 10" 1.05 4.23 x 10" 0.76 Cirumferential Weld Seam 101-171 Notes. (a) Fluence Factor at the 1/4T vessel thickness location. (b) Fluence Factor at the 3/4T vessel thickness location. J Calculation of Adjusted Reference Temperature Revision 1 l l
_. _ -- ~. __ -- . - . . - - = . - - . . 4-11 Contained in Tables 4-10 through 4-13 are the calculations of the ART values used for the generation of the 20 EFPY and 32 EFPY heatup and cooldown cun'es. TABLE 4-10 Calculation of the ART Values for the 1/4T Location @ 20 EFPY Material RG 1.99 R2 CF FF IR h t" ART d Margin ARP Method Intermediate Shell Plate R2005-1 Position 1.1 26.0 F 0.92 -20 F 23.9 F 23.9 F 28*F 2 Intermediate Shell Plate R2005-2 Position 1.1 26.0'F 0.92 -20*F 23.9'F 23.9 F 28*F Intermediate Shell Plate R2005-3 Position 1.1 31.0*F 0.92 -20'F 28.5'F 28.5 F 37'F Lower Shell Plate R2508-1 Position 1.1 58.0*F 0.92 0'F 53.4*F 34*F 87*F Lower Shell Plate R2508-2 Position 1.1 37.0*F 0.92 10*F 34.0'F 34'F 78'F Lower Shell Plate R2508-3 Position 1.1 58.0 F 0.92 40 F 53.4*F 34*F 127 ___-____. --____. - _ . - _ _ _-_______ _ _ _ _ - - --__-__ _ _ _ _'F__ Position 2.1 35.8'F 0.92 40*F 32.9'F 17 F 90 F Intermediate & Lower Shell Position 1.1 31.6'F 0.75 -50*F 23.7*F 23.7 F -3 *F Longitudinal Weld Seam 101-124A & 101-142A Position 2.1 28.3'F 0.75 -50'F 21.2 F 21.2 F -8*F (90* Azimuth) Intermediate & Lower Shell position 1.1 31.6*F 0.92 -50'F 21.9'F 21.9'F 8'F Longitudinal Weld Seam 101-124B,C & 101 142B,C Position 2.1 28.3 F 0.92 -50 F 26.0 F 26.0 F 2F (210' & 330* Azimuth) Intermediate to Lower Shell Position 1.1 31.6*F 0.92 -50 F 29.I 29.1 F 8F ___-.______ _ . _ _ _ _ .- _ _ - - _ _-_______ _ _ - _'F____-_____ __'___ Girth Weld Seam 101-171 Position 2.1 28.3*F 0.92 -50'F 26.0 F 26.0*F 2F Notes, (a) Initial RTmr values are measured values (see Table 4-5). (b) ART = Initial RTer + ARTmyr + Margin ( F) (c) .iRTuor = CF
- FF Calculation of Adjusted Reference Temperature Revision 1
1 4-12 I i i TABLE 4-11 Calculation of the ART Values for the 3/4T Location @ 20 EFPY Material RG 1.99 R2 CF FF 1RTm'd ARTsm.'" Margin ART
- Method Intermediate Shell Plate R2005-1 Position 1.1 26.0*F 0.64 -20 F 16.6'F 16.6 F 13'F Intermediate Shell Plate R2005-2 Position 1.1 26.0*F 0.64 -20*F 16.6*F 16.6*F 13*F Int .rmediate Shell Plate R2005-3 Position 1.1 31.0*F 0.64 -20*F 19.8 F 19.8 F 20*F Lower Shell Plate R2508-1 Position 1.1 58.0*F 0.64 0*F 37.l*F 34.0 F 71 F Lower Shell Plate R2508-2 Position 1.1 37.0 F 0.64 10"F 23.7 F 23.7 F 57'F 40 F 34 F 111 F
__37.1____________ F Lower Shell Plate Pl508-3 Position 1.1 58.0*F 0.64 Position 2.1 35.8 F 0.64 40 F 22.9 F 17 F 80 F Intermediate & Lower Shell Position 1.1 31.6 F 0.49 -50*F 15.5"F 15.5 F -19*F Longitudinal Weld ._ _________ Seam 101 124A & 101-142A Position 2.1 28.3*F 0.49 -50'F 13.9 F 13.9 F -22 F (90 Azimuth) Intermediate & Lower Shell Position 1.1 31.6 F 0.64 -50*F 20.2 F 20.2'F -10 F Longitudinal Weld ______________ _ _ _ _ _ _ ________ _ _ . _ _ _ Seam 101-124B,C & 101-142B,C Position 2.1 28.3'F 0.64 -50 F 18. l'F 18. l'F -14*F (210' & 330' Azimuth) Position 1.1 31.6 F 0.64 -50 F 20.2*F 20.2 F -10 F Intermediate to Lower Shell Girth Weld Seam 101-171 Position 2.1 28.3'F 0.64 -50*F 18. I'F 18. l'F -14*F Notes; (a) Initial RTwr aluesv are measured values (see Table 4-5). (o) ART = Initial RTer + ARTer + Margin ( F) (c) ARTm = CF
- FF Calculation of Adjusted Reference Temperature Revision 1
4-13 TABLE 4-12 Calculation of the ART Values for the 1/4T Location @ 32 EFPY Material RG 1.99 R2 CF FF 1RTer'd ARTn/) Margin ART *) Method (*F) Intermediate Shell Plate R2005-1 Position 1.1 26.0*F 1.05 -20 F 27.3*F 27.3 *F 35'F Intermediate Shell Plate R2005-2 Position 1.1 26.0*F 1.05 -20 F 27.3*F 27.3*F 35*F Intermediate Shell Plate R2005-3 Position 1.1 31.0*F 1.05 -20 F 32.6'F 32.6*F 45*F Lower Shell Plate R2508-1 Position 1.1 58.0*F 1.05 0F 60.9 F 34 F 95 F Lower Shell Plate R2508-2 Position 1.1 37.0'F 1.05 10 F 38.9 F 34*F 83 F Lower Shell Plate R2508-3 Position 1.1 58.0 F 1.05 40 F 60.9*F 34*F 135*F Position 2.1 35.8*F 1.05 40*F 37.6*F 17'F 95'F Intermediate & Lower Shell Position 1.1 31.6*F 0.87 -50 F 27.5 F 27.5'F 5'F Longitudinal Weld Seam 101 124A & 101-142A Position 2.1 28.3*F 0.87 -50*F 24.6*F 24.6*F -l*F (90* Azimuth) Intermediate & Lower Shell Position 1.1 31.6*F 1.05 -50*F 33.2 F 33.2 F 16 F Longi:udinal Weld Seam 101-124B,C & 101 142B,C Position 2.1 28.3'F 1.05 -50 F 29.7*F 28.0*F 8'F (210* & 330* Azimuth) Intermediate to Lower Shell Position 1.1 31.6*F 1.05 -50*F 33.2*F 33.2*F 16*F Girth Weld Seam 101-171 Position 2.1 28.3 F 1.05 -50*F 29.7 F 28.0 F 8*F Notes: (a) Initial RTer values are measured values (see Table 4-5). (b) ART = Initial RTer + ARTer + Margin ( F) (c) ARTer = CF
- FF Calculation of Adjusted Reference Temperature Revision 1 l
~ . -- .. . -.. -- . . - . -
4-14 TABLE 4-13 Calculation of the ART Values for the 3/4T Location @ 32 EFPY - Material RG 1.99 R2 CF FF IkTm" ARTm(4 Margin ART
- Method ('F)
Intermediate Shell Plate R2005-1 Position 1.1 26.0'F 0.76 -20'F 19.8'F 19.8*F 20'F Intermediate Shell Plate R2005-2 Position 1.1 26.0 F 0.76 -20'F 19.8'F 19.8'F 20 F Intermediate Shell Plate R2005-3 Position 1.1 31.0*F 0.76 -20 F 23.6*F 23.6*F 27*F Lower Shell Plate R2508-1 Position 1.1 58.0 F 0.76 0F 44.l'F 34.0 F 78'F Lower Shell Plate R2508-2 Position 1.1 37.0'F 0.76 10 F 28.1 F 28. I'F 66'F l Lower Shell Plate R2508-3 Position 1.1 58.0 F 0.76 40*F 44. l*F 34 118 F __________ _ _ _ ' _ _ .-_____.______________ . _ _ _'F___ _ - - _ _ j Position 2.1 35.8 F 0.76 40 F 27.2'F 17 F 84'F l Intermediate & Lower Shell Position 1.1 31.6*F 0.60 -50 F 19.0 F 19 F -12'F l Longitudinal Weld i Seam 101-124A & 101-142A Position 2.1 28.3 *F 0.60 -50'F 17.0*F 17'F -16*F (90* Azimuth) Intennediate & Lower Shell Position 1.1 31.6*F 0.76 -50 F 24.0*F 24*F -2*F Longitudinal W:Id Scam 101-124B,C & 101-142B,C Position 2.1 28.3'F 0.76 -30*F 21.5'F 21.5'F -7'F (210' & 330' Azimuth) Intermediate to Lower Shell Position 1.1 31.6 F 0.76 -50*F 24.0 F 24*F -2*F Ginh Weld Seam 101-171 Position 2.1 28.3'F 0.76 -50 F 21.5 F 21.5'F -7'F Notes-(a) Initial RTer values are measured values (see Table 4-5). (b) ART = Initial RTer + ARTer + Margin ( F) (c) ARTm = CF
- FF Calculation of Adjust 3d Reference Temperature Revision 1
4-15 The lower shell plate R2508-3 is the limitmg beltline material for all heatur and cooldown cunes to be generated. Contained in Table 4-14 is a summary of the limiting ARTS to be used in the generation of the Wolf Creek reactor vessel heatup and cooldown curves. TABLE 4-14 Summary of the Limiting ART Values Used in the Generation of the Wolf Creek Heatup/Cooldown Cunes EFPY 1/4T Limiting ART 3/4T Limiting ART 20 90 F 80 F 32 95 F 84 F Calculation of Adjusted Reference Temperature Revision 1
5-1 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATG!E LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coohmt system bave been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Section 3 and 4 of this report. This approved methodology is also presented in WCAP-14040-NP-A N , dated January 1996. Figures 5-land 5-3 present the heatup curves with no margins for possible instrumentation errors for a heatup rates of 60 and 100 F/hr. These cunes are applicable for 20 EFPY and 32 EFPY respectively, for the Wolf Creek reactor vessel. Additionally, Figures 5-2 and 5-4 present the cooldown curves with no margins for possible instrumentation errors for cooldown rates of 0,20,40,60, and 100 F/hr. These curves are also applicable for 20 EFPY and 32 EFPY, respectively, for the Wolf Creek reactor vessel. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 5-1 through 5-4. This is in addition to other criteria which must be met before the reactor is made critical, as discussed in the following paragraphs. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 5-1 and 5-3 (for the specific heatup rate being utilized). The straight-line portion of the criticality limit is at the nummum permissible temperature for the 2485 psig insenice hydrostatic test as required by Appendix G to 10 CFR Part 50. The goveming equation for the hydrostatic test is defmed in Appendix G to Section XI of the ASME Codem as follows: I l 1 1.5Kim < Ku (6) where, Kim is the stress intensity factor covered by membrane (pressure) stress, Kr.= 26.78 + 1.233 e p@RN W , T is the nummum permissible metal temperature, and RTer is the metal reference nil-ductility temperature The criticality limit curve specifies pressure-temperature limits for core operation to proside additional margin during actual power production as specified in Reference 2. The pressure-temperature limits for
, core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the muumum temperature required for the insenice hydrostatic test, and at least 40 F higher than the nummum permissible temperature in the corresponding pressure-temperature cun e for heatup and cooldown calculated as described in Section 3 of this report. The vertical line drawn from these points on the pressurc-temperature curve, intersecting a curve 40 F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Heatup and Cooldown Pressure Temper ture Limit Curves Revision 1
k 5-2 Figures 5-1 through 5-4 define all of the above limits for ensuring prevention of nonductile failure for the Wolf Creek reactor vessel. The data points for the heatup and cooldown pressure-temperature limit curves shown in Figures 5-1 through 5-4 are presented in Tables 5-1 through 5-4, respectively. . t 9 l Heatup and Cooldown Pressure Temperature Limit Curves Revision 1
5-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R2508-3 LIMITING ART VALUES AT 20 EFPY: 1/4T,90 F 3/4T,80 F 2500 _ i . ! , , , , , , , , , , , i , , , , i
~ 15s27SSe60 ' i ' + i + i ! / i
- 1. i i i i i i '
! i ! , ! i i !/ i II i i I. '
i i i ! i i +
- .. ! i ! i i ! l i ! i r i
! il . / . V / i li i '
tlltO i . i i i ! i i i 50
- i
* ! MT i !/ / /! i/ i ! i i i i LEAK TEST LIMIT # 'A fl / /! i li i e i i ' ! i g i l' f I 'i !/ i ! i i i
! t t i i i ! i e i i t / / tl / I !/ i a 000 ' ' ' ' ' ' ' ' ' ' '
e
+
i i i v
! ! i i * ! / i / / / i !
! t 1 i !
r i > i e i / ! li / / i i i i i i i
! ! i 6 ! i ! '/ ! ;
i / I !
> i ' ! i ! l l t ! ! ! ! l / 1 ( lI / ! i ! ' i i i i C 1750 i
! I i i 1
UNACCEPTABLE 6
/
/ i/ ill
! /, /i /
/
i 6 . i i i i i i
, I i i i i OPERATION y .' ! 2 1 il l n! / I I ' ! i i i
! ' ' f
/> / !/R5 i i i ! !
g ! I i ! i i i i HEATUP RATE / / i/. \ / i ' ! i ! I ,' i i i to 1500 ! . UP T0 60 F/Hr. i mi "I
/; y6 y
/i/ ! /\ ! .
a i i i i ,, i ! , i F i
; i / ii i W 6
!/i i/ i i t iii i
/i / / i /! \i i C HEATUP RATE /1 i / / i/ _ \ ,
i ! i I i ; i i > i j j i i 1250 i UP TO 100 F/Er. N /i i/ /! / i \! i g 1 , u ,/, z i/1 ,
' \i ACCEPTABLE r*
i i I
! ! i ! ' / N I/ /i ' ! i 0PERATION
~ i i i I /
l /: / /: ! i ! \
! I i i / / I /r /t i ! \ .
1000 / /i / i i + ! '! / i i Ni i i i + i
/ /! i /i,'i i
i ' i i e/r , i t / ' i i/Wii
\' i O
i I i + i / l
/ ! / i/ N i CRIT. LIMIT F0R Z
i i / ;/ /! ! A
\
T 80 F/Hr. L 750 i i i i
, , ij i g
/i /i ;
ri i . N x i i L c i
/
i i , , cmIT. LIMIT i i ! i i i I i i N F0R 100 F/Br. i 6 i 1. . A6 ' i . i i i
! I N W
')
i e i i e i i i i i e 500 ; i
- l ;
l , j ' '
' i O '
i i i } i < Ri i
! i i i i i i ! i i !^vi ! ! i ii e I 6 , i i i i N i i ii i I 250 ' ' ' '
d ' ' ' ' I i ' ' ' ' CRITICALITY LIMIT BASED ON ---
! ! INSERVICE EYDROSTATIC TEST
() ! i - 1i ! i ! ! TJMPERATURE (223 F) FOR TEE i i i __ _ I t i ! i 0 6 1 i i i i i i i l i i i !
.I !
i 3ERYICE PERIOD UP T0 20.0 EFPY a i 0 50 100 150 200 2$0 3d0 3$0 4d0 4$0 500 l Moderator Temperature (Deg.F) FIGURE 5-1 Wolf Creek Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100 F/hr) Applicable to 20 EFPY (Without Margins of for Instrumentation Errors) Heatup and Cooldown Pressure Temperature Limit Curves Revision 1
5-4 MATERIAL PROPERTY BASIS . LIMITING MATERIAL: LOWER SHELL PLATE R2508-3
' \
LIMITING ART VALUES AT 20 EFPY: 1/4T,90 F , 1 3/4T,80 F I 2500 _ i
~ tae37essae i i I 6 i ! ). ! i i i ! ! ! i i i 1 11 i i i i ' ! i i a l !!
e iii, i i i i ! i i ! l ! l ! l > i / ! u 2250 ; i ll ii
';;l i ili
, j!
i i i i ii i 6
, l e
i i i ! i i. l ii l i i i i i I i ! I Ii ' i ! ' I i i ! U3 i i i i i /i i e i i i i i i i i 4 2000 i ! /: I { t ! ! I i e i i
!iii ,
i i
/l l i i !
l'. i i l i, i / ., i i i i- i 1750 O UNACCEPTABLE / i i ! ! ' i OPERATION f ' ' ' i i ' L > i i i i _ c i 'i I ' ' ' u3 1500 - *! > ! ! , ' i i
/ i i j i i u) / ! i i !
. / i i i i i O i i i y 125b , ,, , ,f i
, ACCEPTABLE ! ! i i e
! t i ! / l OPERATION ! i i I 4 ! i! t / i ! ! i I I i i ! If I ! iii i ! !
g 1000 ! 1 i ! .ft i i ! e i !
! I i is ! i i i i e _
. F i COOLDOWN ' ' '
i ! l ' i '
.Wj j "A"8 ' i '
750 - F/Er. ' e _ i, - l l
, i i l
l i ll ii 0 7// /! ! ' '
!i! i l I ! f p ""
*V0-Z:
gn
~
8" 40 t" V! t i
! 8 i i l
i i ! O 60 9'i i l i i i l ' ! i I 6 ! i (( 100 i i i i i i i i ' i i i i t i i ! i i ! ! t 5 250 i '
! i ! I I t l I f f C i ! ' I i I ! i ! !
I i i i i i i _t I i _! r i . t ! ! I i i i i I i i i i 6 e i i i i i 6e i i ! a i i i i 0 50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F) 5 FIGURE 5-2 Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,20,40, 60 and 100*F/hr) Applicable to 20 EFPY (Without Margins for Instrumentation Enors) Heatup and Cooldown Pressure Temperature Limit Curves Revision 1
. _ . . _ _ - _- _ __ __ _ _ _ _ . _ _ . _ . _ ~ _ . . . _ _ _ _ _ . _ _ _
l 5-5 j i i TABLE 5-1 l Wolf Creek Heatup Data at 20 EFPY Without Magins for Instrumentation Errors Run - 1592783850 Heatup Curves 60 F/hr 60 F/hr Cnt. Limit 100 F/hr 100 F/hr Crit. Umit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press. (F) (psi) (F) (psi) (F) (psi) (F) (psf; (psi) (F) 1 60 0 223 0 60 0 223 0 201 2000 ! 60 595 P23 605 -60 567 223 567 223 2485 j 65 596 223 604 65 567 223 567 j 85 596 223 598 85 567 223 567 90 596 223 596 90 567 223 567 l 95 596 223 596 95 567 223 567 l 100 596 223 600 100 567 223 567 105 600 223 606 105 567 223 567 { 110 606 223 615 110 567 223 567 1 115 615 223 626 115 567 223 569 i 120 621 223 639 120 569 223 573 l 125 621 223 654 125 573 223 579 130 621 223 672 130 579 223 587 135 621 223 691 135 587 223 596 140 621 223 712 140 596 223 608 l 140 691 223 736 145 608 223 622 145 712 223 761 150 622 223 637 l l 150 736 223 789 155 637 223 655 155 761 223 820 160 655 223 674 160 789 223 852 165 674 223 696 l 165 820 223 888 170 696 223 719 170 852 223 926 175 719 223 745 175 888 225 967 180 745 225 774 180 926 230 1012 185 774 230 805 185 967 235 1060 190 805 235 838 190 1012 240 1111 195 838 240 875 195 1060 245 1166 200 87C 245 014 200 1111 250 1225 205 914 '250 956 205 1166 255 1289 210 956 255 1002 210 1225 260 1357 215 1002 260 1052 215 1289 265 1430 220 1052 265 1105 l 220 1357 270 1509 225 1105 270 1162 225 1430 275 1593 230 1162 275 1223 230 15C9 280 1663 235 1223 290 1289 235 1593 285 1779 240 1289 235 1360 240 1683 290 1882 245 1360 29'O 1436 245 1779 295 1992 250 1436 295 1517 250 1882 300 2110 255 1517 300 1605 255 1992 305 2235 260 1605 305 1698 260 2110 310 2369 265 1698 310 1798 265 2235 270 1798 315 1905 270 23E9 275 1905 320 2019 280 2019 325 2141 285 2141 330 2271 290 2271 335 2409 295 2409 i i l Heatup and Cooldown Pressure Temperature Limit Curves Revision 1
5-6 l TABLE 5-2 . Wolf Creek Cooldown Data at 20 EFPY
, 'hout Margins for instrumentation Errors Run -1592783850 Cooldown Curves Steady State 20 F/hr 40 F/hr 60 F/hr 100 F/hr .
Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press. (F) (psi) (F) (psi) (F) (psi) (F) (psi) (F) (psi) 60 0 60 0 60 0 60 0 60 0 60 595 60 557 60 518 60 480 60 401 65 605 65 568 65 530 65 492 65 414 70 616 70 579 70 542 70 505 70 429 75 621 75 592 75 556 75 519 75 445 80 621 80 606 80 570 80 534 80 463 85 621 85 620 85 586 85 551 85 482 90 621 90 621 90 602 90 569 90 502 95 621 95 621 95 620 95 588 95 525 100 621 100 621 100 621 100 609 100 549 105 621 105 621 105 621 105 621 105 574 110 621 110 621 110 621 110 621 110 602 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 140 621 140 894 140 876 140 859 140 844 140 822 145 927 145 911 145 897 145 885 145 870 150 962 150 948 150 937 150 929 150 920 155 1000 155 989 155 981 155 975 155 975 160 1040 160 1032 160 1028 160 1026 160 1034 165 1084 165 1079 165 1078 165 1080 170 1130 170 1129 175 1181 180 1234 185 1292 190 1354 195 1421 200 1492 205 1569 210 1651 215 1739 220 1833 225 1934 230 2042 235 2158 240 2281 245 2413 l Heatup and Cooldown Pressure Temperature Limit Curves Revisgn 1
5-7 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R2$08-3 l* LIMITING ARTVALUES AT32 EFPY: 1/4T,95 F 3/4T,84 F 2500 , , , , , , , ,c, , , , , , , , , , , ,
^
[ 1374090780 I i ! ' i i i I/ i ; ! i 1 i . i i i i ! ! i i i
! ! ! i '
i / i ! / '/ 6/ ! / ! ! i ' i i i i i ! i! ! ! ' ' + i ' ' ! . i f ! / /!
" 22r0 1/ /l
! ! i 1 i i i i i a
' i ' '
* =
/ i il / /' i LEAK TEST LIMIT 2 g/ I
! t e i e i +
; i i i <
I /! '
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m '
! / e
/I i/ /! / i i I t i : i . .
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- i i t ! ' / i il l l: /! It i i i e i i i i i ,
a 2000 ' ' i i
' i i !' !' ' ' -
! ' i ! ! '! # ! ! i ! e i / / / i /- e i i! ! i i ! !
v I i i i e i
! i i i ' i a i 6 / / / ! / i i 4 ! ! ! 1 (
)! !
6
!!! i i i i i / / / / i i i i i i '
! i i i t ! ' i r L_ ! i i i f /i i/i i i i i i ' . ! i i !
i e C 1750 ! i r UNACCEPTABLE i i i
/
; i/ i a
/ i/ Y /it i
'I i i ! i i i ! i i i ,
! 4 i . ,
ii y 6 ' OPERATION t / / !/ \ /i i ' ' ! ! i i i
- i i i i ' r /, /. Ji i + i ! i t i ' !
;$ I i i
! i / 1i/ i Al I ! ! i + i * ! ! i i i i m 1500 l l HEATUP RATE i f j
!j#l ! / \!l.
l'l , ll; j ; l,
; i UP TO 60 F/Hr.
, m7 f j j if g, ;ji j j ,i W
i i i I/' / /i iI \ ii! i i ! i i ii i,l((I# O l l HEATUP RATE l fj# fg
' I I U '
1250 ,_, UP TO 100 r/Br. x , , ,f ,j ,, 11 , ACCEPTABLE t .. i! ' M i/ / I/! \i OPERATION ' ! ! Q ! '> il
! iI ' ! ' fi W / /' i I i i i t ' i i t i i i 4
. /t 1 /. / /; ' i .
i I I i 1000 i ' I i i t i i i i i l } e i/ if s/ /
/ i f,
/i
/ {
i i i
\ 4 iii i i ,i j l i T 1 1I . I i I/ t / ,L'- I t i i\
i\ CRIT. LIMIT T 6
' i i O
I i / i /i / h! i i ! > F3R 60 F/Er. I i i i / !/ i i/ N I i i <) a 750 i f /
,/
6 f N i n i _i_ i f, , , , xi , en!T. LIMIT d ' ' ' i/ t ! i i h
/
i
'N N
FOR 100 F/Hr Z a-o I i i u# I i i it i i i! t i
- s 500 l i
l l; l
- l ;;
W
) i i i ' ' ' I i ' i i i
! ! i t !
! ! N I , ! I i ' !
i ! N I 4'4 l t
- ! ! ; . ! ' I i i ! i 250 ' ' ' ' ' ' ' ' ' ' ' ' ' ' '
d ' CRITICALITY LIMIT BASED ON --
' l C) ! !
1 1 ! I! INSERVICE HYDROSTATIC TEST -- i i ' i ! TEMPERATURE (228 F) FOR THE __ t i i i i i
! i i i l ,
i i BERVICE PERIOD UP TO 32.0 IFPY __ 0 0 50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F) FIGURE 5-3 Wolf Creek Reactor Coolant System Heatup Limitations (Heatup Rate of 60 and 100 F/hr) Applicable to 32 EFPY (Without Margins of for Instrumentation Errors) Heatup and Cooldown Pressure Temperature Limit Curves Revision 1
. ~ . . . - - . . - . - . . - . - - . _ _ _ - _ = _ . . -- . - - - . _- -- ~..
5-8 MATERIAL PROPERTY BASIS . LIMITING MATERIAL: LOWER SHELL PLATE R2508-3 LIMITING ART VALUES AT 32 EFPY: 1/4T,95 F 3/4T,84 F 2500 , , , , , , ,, , ,, , ,i., , , , , , , _ 3374sso710 ! i i ! I i! i ! i 6 ! i ! ' ! ! ! ! ! ! iii i l i i J i i ! ' i ! I i i ! i ! i e 4 i i! ! ! ! ! ! ! ! ! I i ! ! i ! ! ! ! $i i li ll ll em 2 2 5 0 ; l ;;l / ; , !,;l l i
,, i i i r i i , i i i
*~
i i i ii i i / i i ei i y) ' i i ! i i i I! i < i ! i l' ,l l /' a 2000 , i
; l ,
i i l i 6 l ll '!/ ' l i l i l i 1 i t ! i O 1750 l'!
! i UNACCEPTABLE 9"7 RATION
[
/!
l l i i l b i
/i i ; i y i
^
i Ii 6 j i 1500 l m . I ! t s i /: t 't ! l ! ,
! ! ! I i /, e i i
U3 l 6i i j / i i I t ' ' / i i t g ! 1250
- y ! ! i ! i / ! i ACCEPTABLE ! i i
' ! ' I ! / i i OPERATION i '! i 4 i ! ' i i !/l 1 i i i i ! ! .I ! i , i 1 i i q
1000 !
' ! ' E i
I ' ! #
! I I;F# 6 i i !
g _ i F , i i COOLDOWN i I/ i i i
" ' ' i i>
750 - mares
' ' I i '
g - F/Hr. .! !
! I i i !
t I 6 0 !i I I A 500 2c " ' ' ' ' ' ' ' ' ' ! 40 4 if I ! ! ! i ' ' ! O __ 60 F i ! i ! i 1 5
-- "' 'l '
l i; l l l i 250 , ; l , , i , ; , , l , iii ta i , i i , 4 ; i i ! i ii i i t i
' ' i
!i o
0 50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F) < FIGURE 5-4 Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,20,40, 60 and 100*F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors) i Heatup and Cooldown Pressure Temperature Limit Curves ReviSiO" 1
5-9 TABLE 5-3 Wolf Creek Heatup Data at 32 EFPY Without Margins for Instrumentation Errors Run = 1374690710 Heatup Curves 60 F/hr 60 F/hr Crit. Limit 100 F/hr 100 F/hr Crit. Umit Leak Test l'imit Temp. Pres. Temp. Pres. Temp. Pres. Temp. Pres. Temp. Pres. (F) (psi) (F) (psi) (F) (psi) (F) (psi) (F) (psi) 60 0 228 0 60 0 228 0 206- 2000 60 584 228 595 60 554 228 554 228 2485 l 65 584 228 594 65 554 228 554 85 584 228 Se7 85 554 228 554 90 584 228 584 90 554 228 554 95 584 228 585 95 554 228 554 100 585 228 588 100 554 228 554 105 588 228 593 105 554 228 554 110 593 228 601 110 554 228 554 115 601 228 611 115 554 228 556 120 611 228 624 120 556 228 559 125 621 928 638 125 559 228 564 130 621 228 654 130 564 228 572 135 621 228 672 135 572 228 581 140 621 228 692 -140 581 228 591 140 672 228 714 145 591 228 604 145 692 228 738 150 604 228 618 150 714 228 764 155 618 228 635 155 738 228 793 160 635 228 653 160 764 228 824 165 653 228 673 165 793 228 857 170 673 228 695 170 824 228 893 175 695 228 719 175 857 228 932 180 719 228 746 180 893 230 974 185 746 230 775 185 932 235 1019 190 775 235 807 i 190 974 240 1068 195 807 240 841 195 1019 245 1120 200 841 245 878 200 '1068 250 1176 205 878 250 918 205 1120- 255 1236 210 918 255 961 210 1176 260 1300 215 961 260 1008 21 5 1236 265 1369 220 1008 265 1058 220 1300 270 1444 225 1058 270 1112 225 1369 275 1523 230 1112 275 1170 230 1444 280 1608 235 1170 280 1232 235 .1523 285 1699 240 1232 285 1299 240 1608 290 1797 245 1299 290 1371 245 1699 295 1901 250 1371 295 1448 250 1797 300 2013 255 1448 300 1530 255 1901 305 2132 260 1530 305 1619 260 2013 310 2259 265 1619 310 1713 265 2132 315 2394 270 1713 315 1814 270 2259 275 1814 320 1923 275 2394 280 1923 325 2038 285 2038 330 2161 r 290 2161 335 2293 l 295 2293 340 2433 300 2433 Heatup and Cooldown Pressure Temperature Limit Curves Revision 1
5-10 TABLE 5-4 Wolf Creek Cooldown Data at 32 EFPY Without Margins for Instnunentation Errors Run - 1374690710 Cooldown Curves Steady State 20 F/hr 40 F/hr 60 F/nr 100 F/hr Temp. Pres. Temp. Pres. Temp. Pres. Temp. Pres. Temp. Pres. (F) (psi) (F) (psi) (F) (psi) (F) (psi) (F) (psi) 60 0 60 0 60 0 60 0 60 0 60 585 60 547 60 507 60 468 60 387 65 595 65 557 65 518 65 479 65 400 70 605 70 568 70 530 70 491 70 413 75 616 75 579 75 542 75 504 75 428 80 621 80 592 80 555 80 519 80 445 85 621 85 605 85 570 85 534 85 462 90 621 90 620 90 585 90 550 90 481 95 621 95 621 95 602 95 568 95 502 100 621 100 621 100 C20 100 588 100 524 105 621 105 621 105 621 105 608 105 548 110 621 110 621 110 621 110 621 110 574 115 621 116 621 115 621 115 621 115 602 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 130 621 -130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 140 621 140 864 140 843 140 824 140 806 140 778 145 894 145 876 145 859 145 844 145 822 150 927 150 911 150 896 150 885 150 869 ? 155 962 155 948 155 937 155 928 155 920 $ 160 1000 160 989 ' 160 981 160 975 160 975 165 1040 165 1032 165 1027 165 1026 165 1034 170 1084 170 1079 170 1078 170 1080 175 1130 175 1129 180 1181 185 1234 190 1292 195 1354 200 1421 205 1492 21 0 1569 21 5 1651 220 1739 225 1833 230 1934 235 2042 ' 240 2158 245 2281 250- 2413 Heatup and Cooldown Pressure Temperature Limit Curves Revision 1
I 6-1
- 6 REFERENCES I Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S. l Nuclear Regulatory Commission, May,1988.
2 10 CFR Part 50, Appendix Q " Fracture Toughness Requirements", Federal Register, Volume 60, No. 243, dated December 19,1995. 3 1989 ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Appendix Q " Fracture Toughness Criteria for Protection Against Failure". 4 CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1996. 5 WCAP-13365, Rev.1, " Analysis of Capsule Y from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program", J.M. Chicots, et al., April 1993, 6 1989 Section III, Division 1 of the ASME Boiler and Pressure %ssel Code, Paragraph NB-2331,
" Material for Vessels".
7 WCAP-15078, " Analysis of Capsule V from the Wolf Creek Nuclear Operating Corporation Wolf Creek Reactor Vessel Radiation Surveillance Program", Ed Terek, et al., August 1998. 8 WCAP-14040-NP-A, Revision 2, " Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D. Andrachek, et al., January 1996. 9 CE NPSD-1039, Rev. 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds, Appendix A, CE Reactor Vessel Weld Properties Database, Volume 1," CEOG Task 902, June 1997. References Revision 1}}