ML20211C220
| ML20211C220 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 07/31/1999 |
| From: | Christopher Boyd, Terek E, Trombola D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20211C211 | List: |
| References | |
| WCAP-15102, WCAP-15102-R01, NUDOCS 9908250179 | |
| Download: ML20211C220 (47) | |
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+ ++++++++++ WCAP - 15102 iu,.isi,,,, ,
/
V. C. Summer Unit 1
. H.eatup and Cooldown Limit Curves for Normal Operation
. l Westinghouse Electric Company LLC W. em. -
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15102, Revision 1 V. C. Summer Unit 1 Heatup and Cooldown Limit Curves for Normal Operation Ed Terek July 1999 Work Performed Under Shop Order CJWP139Q Prepared by the Westinghouse Electric Company, LLC for the South Carolina Electric and Gas Company e/
Approved:
D. M. Trombola, Manager Mechanical Systems Integration
. Approved: '
m C.%
C. H. Boyd, Manager Equipment & Materials Technology Westinghouse Electric Company LLC Energy Sy, stems P.O. Box 355 Pittsburgh, PA 15230-0355 C1999 Westinghouse Electric Company LLC All Rights Reserved
iii TABLE OF CONTENTS LIST OF TABLES ... . . . ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .... . .. iv LIST OF FIGURES... .... . . . . . . . . . . . . . . . . . . .. .. . . . . . . . . . . ..vi PREFACE . .... .. . . .. . . . . . . . . . . .. . . . . . . . . . . ... .. . . . . vii EXECUTIVE
SUMMARY
. . ... . . . . . . . . . . . . . . . . . . . . . .. . . . . viii 1 INTRODUCTION..... .. . . . . .. .. . . . . . . ... ... . . . . . . .
2 BACKGROUND AND PURPOSE.... . . . . . .. . ..... . 2- 1 3
CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS . . . . . 3-1 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE.. . . . . . . . . 4-1 5
HEA FUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES.. . . . . .. 5-1 6 REFERENCES. . ... . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 6- 1 Revision 1
iv LIST OF TABLES Table 4-1 Summary of the Peak Pressure Vessel Neutron Fluence Values at 20 EFPY used for the calculation of ART Values (n/cm2, 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 of ART Values (n/cm2, E > 1.0 MeV).. . . . .4-3 Table 4-3 Calculated Integrated Neutron Exposure of the V. C. Summer Unit 1 Surveillance Capsules Tested to Date.. . . . . .. . . . . . .4-4 Table 4-4 Measured 30 ft-lb Transition Temp;rature Shifts of the Beltline Materials
- Contained in the Surveillance Program.. . . . . .4-5 e
Table 4-5 Reactor Vessel Beltline Material Unitradiated Toughness Properties.. . . 4-6 Table 4-6 Calculation of Chemistry Factors using V. C. Summer Unit 1 Surveillance Capsule Data .. .. .. . ... . . . . . .. .. .4-7 Table 4-7 Summary of the V. C. Summer Unit 1 Reactor Vessel Beltline Material Chemistry Factors Based on Regulatcry Guide 1.99, Revision 2, Position 1.1 and Position 2.1.. 4-8 Table 4-8 Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 20 EFPY Heatup and Cooldown Curves... . . .. .. 4-9 Table 4-9 Calculation of the 1/4T and 3/4T Fluence Factors Values used for the Generation of the 32 EFPY Heatup and Cooldown Curves.. .. . . . . . . . .4-10 Table 4-10 .4-11 Calculation of the ART Values for the 1/4T Location @ 20 EFPY.. .
Table 4-11 .. 4-12 Calculation of the ART Values for the 3/4T Location @ 20 EFPY.. .
Table 4-12 .4-13 Calculation of the ART Values for the 1/4T Location @ 32 EFPY.. . ..
Table 4-13 Calculation of the ART Values for the 3/4T Location @ 32 EFPY.. . .4-14 Table 4-14 Summary of the Limiting ART Values used in the Generation of the V. C. Summer Unit 1 Heatup/Cooldown Curves. ... . . . . . . .. . 4-15 Table 5-1 V. C. Summer Unit 1 Heatup Data at 20 EFPY (Without Margins for Instrumentation Errors).. . . . .. .. .. .
. 5-5 Table 5-2 V. C. Summer Unit 1 Cooldown Data at 20 EFPY (Without Margins for Instrumeritation Errors).. . .. . . . . . . . . 5-6 Revision 1
v Table 5-3 V. C. Summer Unit 1 Heatup Data at 32 EFPY (Without Margins for Instrumentation Errors).. .. . .. . . 5-9 Table 5-4 V. C. Summer Unit 1 Cooldown Data at 32 EFPY (Without Margins for Instrumentation Errors).. . . . . . 5-10 b
a 4
Revision 1
vi LIST OF FIGURES Figure 3-1 Geometry of the Upper Head / Flange Region of a Typical Westinghouse Four Loop Plant Reactor Vessel . .. .. .. ... . . . . .. ..... . . . . . . . . . . . . .. . ... . . 3-9 Figure 3-2 Crack Driving Force as a Function of Flaw Size: Outside Surface Flaw in the Closure Head to Flange Region Weld ... . . ... ... . .. ... .. .. . .... .. . . . . . ..... . 3-10 Figure 3 3 Determination of Boltup Requirement, using KIC........ .. .... ..... .... 3-11
. Figure 51 V. C. Summer Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 50 and 100 F/hr) Applicable to 20 EFPY (Without Margins for Instrumentation Errors)... . . . . . ..... ... ... . ..... ......... 5-3 Figure 5-2 V. C. Summer Unit i Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100'F/hr) Applicable to 20 EFPY (Without Margins for Instrumentation Errors)..... . ... ... ..... . ................5-4 Figure 5-3 V. C. Summer Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 50 and 100 F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors)...... ...... .. .. ... . ... .. ... . . . . ... ..... 5-7 Figure 5-4 V. C. Summer Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100 F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors).: . . . . . . . . . . . . . . . . . .58 4
Revision 1
vii PREFACE This report has been technically reviewed and verified by:
Reviewer: T. J. Laubham .
/V b
G 9
e 4
s 5
Revision 1
viii EXECUTIVE
SUMMARY
Revision 0:
The purpose of this report is to generate pressure-temperature limit curves for V. C. Summer Unit I for normal operation at 20 and 32 EFPY using the methodology from WCAP 14040-NP-A which encompasses the requirements of the 1989 ASME Boiler and Pressure Vessel Code,Section XI, Appendix G. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) 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 lower shell plates C9923-1 and C9923-2 (i.e. the limiting beltline
. region material). The pressure-temperature limit curves were generated for heatup rates of 50 and 100 F/hr and cooldown rates of 0,25,50 and 100 F/hr. These curves can be found in Figures 5-1 through 5-4.
Revision 1:
The purpose of this revision is to generate pressure pressure-temperature limit curves for V. C. Summer Unit I for normal operation at 20 and 32 EFPY utilizing updated methodology and without instrumentation error margins. The updated methodology that is being employed is the utilization of the 1996 ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, along with ASME Code Case N-640. In addition, this report contains ajustification for a reduced flange temperature requirement which was also incorporated into the pressure-temperature curves provided in this report. All other calculations / data remains unchanged.
r O
Revision i
1-1 1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTNDT(reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RTNDT (ART) 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 ARTNDT, and adding a margin. The unirradiated RTNDT si designated as the higher of either the drop weight nil-ductility transition temperature (NDTF) or the temperature at which the material exhibits at least 50 ft-lb ofimpact energy and 35-mil lateral expansion (transverse to the major rolling direction for late material) minus 60 F.
RTNDT ncreases i as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT ue d to the radiation exposure associated with that time period must be added to the unirradiated/ initial RTNDT(IRTNDT). The extent of the shift in RTNDT si 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"[Il. Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTNDT + ARTNDT + margins for uncertainties) at the %T and
%T locations, where T is the thickness of the vessel at the beltline region measured from the clad / base metal interface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves for normal operation.
6 m
l Revision 1
2-1 2 BACKGROUND AND PURPOSE Appendix G to the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1," Fracture Toughness Criteria for Protection Against Failure"[3] was updated in 1996 and ASME Code Case N-640,
" Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1"[4), was approved in February of 1999. The 1996 ASME Section XI, Appendix G, provides a more accurate methodology for calculating stress intensity factors due to the thermal and pressure stresses at the %T and %T locations while Code Case N-640 allows the use of the KIC methodology rather than the K1 A methodology.
In September of 1998 Westinghouse completed the analysis surveillance capsule W from the V. C.
Summer Unit I reactor vessel. As a part of this analysis Westinghouse generated new heatup and cooldown curves for 20 and S2 EFPY. The heatup and cooldown curves were developed per the methodology given in WCAP-14040-NP-A, Revision 2," Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown L,imit Curves"[5] and included generic margins for instrumentation errors. In Addition, these curves included 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[2],
The purpose of this revision is to present the calculations and development of the South Carolina Electric
& Gas Company V. C. Summer Unit I pressure-temperature curves for 20 and 32 EFPY utilizing the 1996 Appendix G to the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1,
" Fracture Toughness Criteria for Protection Against Failure"along with ASME Code Case N-640.
" Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1" In addition, this report provides technicaljustification for relaxing the flange temperature requirement of Appendix G to 10 CFR Part 50 based on the use of the KlC methodology rather than the Kl A methodology. These pressure-temperature curves are being developed for normal operation up to 20 and 32 EFPY and do not include margins for instrumentation errors.
This report documents the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2[1], for all the beltline materials and the development of the heatup and cooldown pressure-temperature limit curves for normal operation.
9 Revision 1
3-1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 Overall Approach a
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 ofligh: water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements.Section XI, Division 1, " Rules for Inservice inspection of Nuclear Power Plant Components", Appendix G[3], 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 ,1 for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kl e , for the metal temperature at that time. K le is obtained from the reference fracture toughness curve, defined in Code Case N-640 of ASME Appendix G to Section XI. The KI c cun e is given by the following equation:
L = 332 + 20.734
- etu2< r-nren (;)
where, K Ic = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K el curve is based on the lower bound of static critical K 1values measured as a function of
, temperature on specimens of SA 533 Grade B Class 1, SA-508-2, and SA-508-3 steels.
Revision 1
3-2 3.2 Methodology for Pressure-Temperature Limit Curve Development The governing equation for the heatup-cooldown analysis is defined as:
C
- Ki.+ Kn < Kic (2)
~
where, K im = stress intensity factor caused by membrane (pressure) stress
. kit = stress intensity factor caused by the thermal gradients KlC = function of temperature relative to the RTNDT 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 For membrane tension, the Kg corresponding to membrane tension for the postulated defect is:
K Im = Mm * (PRi + t) (3)
Where Mm for an inside surface is given by:
Mm = 1.85 for Vt < 2, Mm = 0.926 Vt for 2 s Vt s 3.464, and '
Mm = 3.21 for Vt > 3.464.
Similarly, Mm for an outside surface flaw is given by:
Mm = 1.77 for Yt < 2, Mm = 0.893 Vt for 2 s Vt s 3.464, and
. Mm = 3.09 for Vt > 3.464.
where:
p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.
Revision 1
s 3-3 For Bending Stress, the KI corresponding to bending stress for the postulated defcet is:
K ib = M b* maximum bending stress, where Mb is two-thirds of Mm For the Radial Thermal Gradient, the maximum KI produced by radial thermal gradient for the postulated inside surface defect is:
K It= 0.953x10-3 x CR x 12 5 (4) where:
CR = the cooldown rate in 'F&r.
For the Radial Thermal Gradient, the maximum KI produced by radial thermal gradient for the postulated outside surface defect is:
KIt= 0.753x10-3 x HU x 12 5 (5) where:
HU = the heatup rate in 'F/hr.
The through wall temperature difference associated with the maximum thermal K1 can be determined from ASME Section XI,' Appendix G, Figure G-2214-1. The temperature at any radial distance from the
- 2 vessel surface can be determined from ASME Section XI, Appendix G, Figure G-2214-2 for the maximum thermal KI .
(a) The maximum thermal K 1relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(1) and (2) of Appendix G to ASME Section XI.
Revision 1
34 (b) Alternatively, the K for 1 radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a %-thickness inside surface defect using the relationship:
Ki, = (1.0359Co + 0.6322Ci + 0.4753C2 + 0.3855C3)
- 6 (6)
I or similarly, K IT uring d heatup for a %-thickness outside surface defect using the relationship:
1 l
Ki, = (1.043Co + 0.630C + 0.481C2 + 0.401C3)
- 6 (7) 1 where the coefficients Co, C1 , C2and C 3are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:
cr(x) = Co + Ci(x / a) + C2(x / a)' + C3(x / a)' (8) l l
and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.
Note, that equations 3 through 8 were added to the OPERLIM computer program, which is the Westinghouse computer program used to generate pressure-temperature limit curves. No other changes were made to the OPERLIM computer program with regard to the pressure-temperature curve calculation
, methodology. Hence, the pressure-temperature curve methodology described in WCAP-14040[5]
Section 2.6 (equations 2.6.2-4 and 2.6.3-1) remains valid for the generation of the pressure-temperature curves documented in this report with the exceptions described above.
l i
i Revision 1 j
3-5 At any time during the heatup or cooldown transient, KlC si determined by the metal temperature at the tip of a postulated flaw at the %T and %T location, the appropriate value for RTNDT, 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 It , 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 of Appendix 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 of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procediare 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 %T 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 DT (temperature) developed during cooldown results in a higher value of KlC at the
%T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K lC exceeds KIt, the calculated allowable pressure during cooldown will be greater I than the steady-state value.
The above procedures are needed because there is no direct control on temperature at the %T 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 ]
cnsures conservative operation of the system for the entire cooldown period.
I 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 hestup rate conditions assuming the presence of a %T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses
- l. )
produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K 1C or f the %T crack during heatup is lower than the K lC or f the %T 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 KlC values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the %T flaw is considered.
Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower i
Revision 1
3-6 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 %T flaw located at the %T 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 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 individual basis.
Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a l point-by-point comparison of the steady-state and finite heatup rate data. At eny given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under considgration.
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.
Revision 1
3-7 3.3 Closure Head / Vessel Flange Requirements 10 CFR Part 50, Appendix G contains the requirements for the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTNDT yb at least 120*F for normal operation when the pressure exceeds 20 percent of the pre-service hydrostatic test pressure (3106 psig), which is 621 psig for the V.
C. Summer Unit I reactor vessel.
This requirement was originally based on concerns about the fracture margin in the closure flange region.
During the boltup process, stresses in this region typically reach over 70 percent of the steady-state stress.
without being at steady-state temperature. The margin of 120F and the pressure limitation of 20 percent of hydrotest pressure were developed using the K ia fracture toughness, in the mid 1970s.
Improved knowledge of fracture toughness and other issues which affect the integrity of the reactor j vessel have led to the recent change to allow the use of K ic in the development of pressure-temperature l ~
curves, as contained in Code Case N-640," Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1". The following discussion uses a similar approach (i.e. using K ic) is used here to develop equivalent flange requirements.
l The geometry of the closure head flange region for a typical Westinghouse four loop plant reactor vessel !
which is more conservative than the geometry of a three loop plant reactor vessel such as the V. C.
Summer Unit I reactor vessel is shown in Figure 3-1. The stresses in this region are highest near the outside surface of the head. Hence, a outside reference flaw of 25 percent of the wall thickness parallel to the dome to flange weld (i.e. in the direction of welding) was postulated in this region. To be consistent with ASME Section XI, Appendix G, a safety factor of two was applied and a fracture calculation performed.
Figure 3-2 shows the crack driving force or stress intensity factor for the postulated flaw in this region, along with a second curve which incorporates the safety factor of two. Note that the stress intensity factor with a safety factor of one for this region does not exceed 55 ksiVin., even for postulated flaws of up to 50 percent of the wall thickness. For the reference flaw, with the safety factor of two, the applied stress intensity factor is 85.15 ksiVin. at 25 percent of the wall thickness.
The determination of the bolt-up, or flange requirement, is shown in Figure 3-3, where the fracture toughness is plotted as a function of the temperature. In this figure, the intersection between the stress intensity factor curve and the Kia toughness curve occurs at a value slightly higher than
. T-RTNDT = 100*F, which is in the range of the existing 120 F requirement. The reference calculation used for the original requirement (which is no longer available) resulted in a temperature requirement of T-RTNDT = 120 F. This corresponds to a K ia (with a safety factor of 2) of 98 ksiVin. Note that the use '
of the K ei curve to determine this requirement results in a revised requirement of T-RTNDT = 45 F, as seen in Figure 3-3.
j i
Revision 1 l
ou., m 3-8 Therefore, the appropriate flange requirement for use with the Kje curve is as follows:
The pressure in the vessel should not exceed 20 percent of the pre-service hydro-test pressure until the temperature exceeds T- RTNDT = 45'F. This requirement has been implemented with the curves presented in this report.
The limiting unirradiated RTNDT of 10 F (Table 4-5 in of this report) occurs in closure head flange 5297-V1 of the V. C. Summer Unit I reactor vessel, so the minimum allowable temperature of this region is 55'F at pressures greater than 621 psig with no margin for uncertainties.
d l
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)
3-9 205.00 DI A =
191.875 DI A +
1 TOP HEAD DOME TORUS
+/ - 'TO FLANGE WELD
~
n' NOTE: ALL DIMENSIONS ARE IN INCH'ES 27.25 167.00 DI A l g 172.56 DI A z n h g w 18.0 167.00 DI A o 27.625 VESSEL FLANGE TO \ U mTE: DIMENSIONS DO NOT UPPER SHELL WELD-INCLUDE CLADDING 170.88 DI A r +
+ e UPPER HEAD REGION Figure 3-1 Geometry of the Upper Head / Flange Region of a Typical Westinghouse Four Loop Plant Reactor Vessel Revision 1
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QCD Q3 Q10 Q15 Q2D Q25 QE Q35 00 Q45 Q!D Q2 Q80 085 070 nunmenmenemmanemonen Figure 3-2 Crack Driving Force as a Function of Flaw Size: Outside Surface Flaw in the Closure Head to Flange Region Weld Revision 1
1 3-11 a
m 4
D v
l 150 D
i f.
ju t: -
m remar]
0 g g 150 E #
TKndt(thS O Figure 3-3 Determination of Boltup Requirement, using Kei Revision i
1 l
4-1 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART = InitialRTnr + A RTwr + Margin (9y Initial RTNDT si the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [6]. If measured values ofinitial RTNDT 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. -
ARTNDT si the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows:
A RTer = CF *l'"'*"" (l0)
To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
[rargw = fog,
- e### (l1) where x inches (vessel beltline thickness is 7.75 inchesl7}) is the depth into the vessel wall measured from the vessel clad / base metal interface. The resultant fluence is then placed in Equation 4 to calculate the ARTNDT at the specific depth.
The Westinghouse Radiation Engineering and Analysis group evaluated the vessel fluence projections and the results are presented in Section 6 of WCAP-1510l[8] . 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"[5). Tables 4-1 and 4-2, herein, contain the calculated vessel surface fluence values along with the Regulatory Guide 1.99, Revision 2,1/4T and 3/4T calculated fluences used to calculate the ART values for all beltline materials in the V. C. Summer Unit I reactor vessel.
Additionally, the calculated surveillance capsule fluence values are presented in Table 4-3.
O Revision 1 i
4-2 TABLE 4-1 Summary of the Peak Pressure Vessel Neutron Fluence Values at 20 EFPY used for the Calculation of ART Values (n/cm2, E > 1.0 MeV)
Azimuth Surface %T %T Intermediate Shell Plate A9154-1 2.45 x 1019 1.54 x 1019 6.07 x 1018 Intermediate Shell Plate A9153 2 2.45 x 1019 1.54 x 1019 6.07 x 1018 Lower Shell Plate C9923-1 2.45 x 1019 1.54 x 1019 6.07 x 1018 Lower Shell Plate C9923-2 2.45 x 1019 1.54 x 1019 6.07 x 1018 Intermediate Lower Shell Longitudinal 0.892 x 1019 5.60 x 1018 2.21 x 1018 Weld Seam BC, BD and BA, BB (0. Azimuth)
Intermediate to Lower Shell 2.45 x 1019 1.54 x 1019 6.07 x 1018 Cirumferential Weld Seam AB O
Revision 1
1 4-3 TABLE 4-2 Summary of the Peak Pressure Vessel Neutron Fluence Values at 32 EFPY used for the Calculation of ART Values (n/cm2, E > 1.0 MeV)
Material Surface %T %T Intermediate Shell Plate A9154-1 3,g4 x 1019 2.41 x 1019 9.52 x 1018 Intermediate Shell Plate A9153-1 3.84 x 1019 2.41 x 1019 9.52 x 1018 Lower Shell Plate C9923-1 3.84 x 1019 2.41 x 1019
. 9.52 x 1018 Lower Shell Plate C9923-2 3.84 x 1019 2.41 x 1019 9.52 x 1018 Intermediate & Lower Shell Longitudinal 1.43 x 1019 8.98 x 1018 3.54 x 1018 Weld Seam BC, BD and BA, BB (45' Azimuth)
Intermediate to Lower Shell 3,g4 x 1919 2.41 x 1019 9.52 x 1018 Cirumferential Weld Seam AB o
Revision 1
4-4 TABLE 4-3 Calculated Integrated Neutron Exposure of the V. C. Summer Unit 1 Surveillance Capsules Tested to Date Capsule Fluence U 6.542 x 1018 n/cm2,(E > 1.0 MeV)
V 1.538 x 1019 n/cm2,(E > 1.0 MeV)
X 2.543 x 1019 n/cm2,(E > 1.0 meV)
~
W 4.664 x 1019 n/cm2,(E > 1.0 MeV) 2 Margin is calculated as, M = 2da' + a . The standard deviation for the initial RTNDT margin term, oi, is 0*F when the initial RTNDT si a measured value, and 17 F when a generic value is used. The standard deviation for the ARTNDT margin term, og, is 17 F for plates when surveillance capsule data is not used and 8.5'F for plates when surveillance capsule data is used. For welds, og is 28'F when surveillance capsule data is not used and 14'F when surveillance capsule data is used. In addition, og need not exceed one-half the mean value of ARTNDT-O Revision 1
4-5 Contained in Table 4-4 is a summary of the Measured 30 ft-lb transition temperature shifts of the beltline materialsl8). These measured shift values were obtained using CVGRAPH, Version 4.ll9], which is a hyperbolic tangent curve-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-Ib Transition
~
Temperature Shift (a)
, Intermediate Shell Plate A9154-1 U 36.0 F V 52.6 F X 37.7 F (Longitudinal Orientation) W 65.7 F Intermediate Shell Plate A9154-1 U 14.5 F V 32.4 F X 26.0 F (Transverse Orientation) W 57.8 F Surveillance Program U 22.2'F V 46.5 F Weld Metal X 22.4'F W 43.3'F U 35.6 F
. V 50.1 F Heat Affected Zone X 54.1 F
~
W 60.3 F Notes:
(a) From capsule W dosimetry analysis results[8],(x1019 n/cm2, E>l.0.MeV).
Revision 1
4-6 Table 4-5 contains a summary of the weight percent of copper, the weight percent of nickel and the initial RTNDT of the 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 Unitradiated Toughness Properties [7 & 10]
Material Description Cu(%) Ni(%) Initial RTNDT(a)
Closure Head Flange 5297-V1(b) .. .. Icop(b)
Vessel Flange 5301-V-1 -- --
0 F(b)
Intermediate Shell Plate A9154-1 0.10 0.51 30 F Intermediate Shell Plate A9153-2 0.09 0.45 -20 F Lower Shell Plate C9923-1 0.08 0.41 10 F Lower Shell Plate C9923-2 0.08 0.41 10 F Intermediate Shell Longitudinal Welds 0.05 0.91 -44 F Seams BC & BD Intermediate Shell Longitudinal Welds 0.05 0.91 -44'F Seams BA & BB Intermediate Lower Shell Plate 0.05 0.91 -44 F Circumferential Weld Seam AS Surveillance Program Weld Metal 0.04 0.95 ---
Notes:
. (a) The initial RTm values for the plates and welds are based on measured data per WCAP-12867t'l (b) In the past the closure head flange was reported as Heat A9231 with a IRTm of-20*F. Based on a review of Westinghouse files, the correct data is Heat # 5297-VI with a IRTm of 10'F. Also, the vessel flange was reported a IRTm of 10*F., however, based on a review Westinghouse files, the correct IRT, of 0*F.
Revision 1
4-7 TABLE 4-6 Calculation of Chemistry Factors using V. C. Summer Unit 1 Surveillance Capsule Data Material Capsule Capsule f* FF* ARTY,/4 FF* ART,or FF 8 Intermediate Shell U 0.654 0.881 36.0 31.7 0776 Plate A9154-1 V 1.538 1.119 52.6 58.9 1.252 X 2.543 1.250 37.7 47.1 1.563 (Longitudinal) W 4.664 1.388 65.7 91.2 1.927 o intermediate Shell U 0.654 0.881 14.5 12.8 0.776 Plate A9154-1 V 1.538 1.119 32.4 36.3 1.252 X 2.543 1.250 26.0 32.5 1.563 (Transverse) W 4.664 1.388 57.8 80.2 1.927 SUM 390.7 11.036 CFA9154-1 = E(FF
- RTNDT) + E(FF2) = (390.7) + (ll.036) = 35.4 F Surveillance Weld U 0.654 0.881 28.0(d) 24.7 0776 Material V 1.538 1.119 58.6(d) 65.6 1.252 X 2.543 1.250 28.3(d) 35.4 1.563 W 4.664 1.388 54.4(d) 75.5 1.927 SUM 201.2 5.518 CFWeld = E(FF
(a) F= Measured fluence from capsule W dosimetry analysis results(8)(x 1019 n/cm2, E > 1.0
. MeV).
(b) FF = fluence factor = f(0.28 - 0.l
- log f)
(c) RTNDT values are the measured 30 ft-lb shift values.
(d) The surveillance weld metal ARTNDT values have been adjusted by a ratio of 1.26.
(CFyw + CFSW = 68 F + 54*F = 1.26)
Revision 1
4-8 TABLE 4-7 Summary of the V. C. Summer Unit 1 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 A9154-1 65.0 F 35.4 F Intermediate Shell Plate A9153-2 58.0 F ---
Lower Shell Plate C9923-1 51.0 F --
Lower Shell Plate C9923-2 51.0 F ---
Intermediate Shell Longitudinal Weld 68.0 F 36.5 F Scams BC & BD Lower Shell Longitudinal Weld Seams, 68.0 F 36.5 F BA & BB Intermediate to Lower Shell Plate 68.0 F 36.5'F Circumferential Weld Seam AB Note: See Reference 8 for the credibility evaluation of the V.C. Summer Unit I surveillance data.
ed' s l
Revision 1 l
4-9 Contained in Tables 4-8 and 4-9 are summaries of the fluence factors (FF) used in the calculation of adjusted reference temperatures for the V. C. Summer Unit I reactor vessel beltline materials for 20 EFPY and 32 EFPY TABLE 4-8 Calculation of the 1/4T and 3/4 T Fluence factor Values used for the Generation of the 20 EPFY Heatup/Cooldown Curves
, Azimuth 1/4 T F 1/4T FF 3/4T F 3/4 T FF (n/cm2 E > I.0 MeV) (n/cm2, E >l.0 MeV)
Intermediate Shell Plate A9154-1 1.54 x 1019 1.12 6.07 x 1018 0.860 Intermediate Shell Plate A9153-2 1.54 x 1019 1.12 6.07 x 1018 0.860 Lower Shell Plate C9923-1 1.54 x 1019 1.12 6.07 x 1018 0.860 Lower Shell Plate C9923-2 1.54 x 1019 1.12 6.07 x 1018 0.860 Intermediate & Lower Shell 5.60 x 1018 0.838 2.21 x 1018 0.594 Longitudinal Weld Seams BC, BD and BA, BB (45" Azimuth)
Intermediate to Lower Shell 1.54 x 1019 1.12 6.07 x 1019 0.860 Cirumferential Weld Seam AB 4
Revision 1
1 1
4-10 TABLE 4-9 Calculation of the 1/4T and 3/4 T Fluence Factor Values used for the Generation of the 32 EPFY Heatup/Cooldown Curves Azimuth 1/4 T F 1/4T FF 3/4T F 3/4 T FF (n/cm2 E > 1.0 MeV) (n/cm2. E >l.0 MeV)
Intermediate Shell Plate A9154-1 2.41 x 1019 1.24 9.52 x 1018 0.986 Intermediate Shell Plate A9153-2 2.41 x 1019 1.24 9.52 x 1018 0.986 Lower Shell Plate C9923-1 2.41 x 1019 1.24 9.52 x 1018 0.986 Lower Shell Plate C9923-2 2.41 x 1019 1.24 9.52 x 1018 0.986 Intermediate & Lower Shell 8.98 x 1018 0.970 3.54 x 1018 0.713 Longitudinal Weld Seams BC, BD and BA, BB (45" Azimuth)
Intermediate to Lower Shell 2.41 x 1019 1.24 9.52 x 1018 0.986 Cirumferential Weld Seam A5 dr Revision 1 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 curves.
TABLE 4-10 Calculation of the ART Values for the 1/4T Location @ 20 EFPY
- Material RG 1.99 R2 CF FF 1RT,,/*8 ART,,/o Margin ART
- Method ('F)
Intermediate Shell Plate A91541 Position 2.1 35.4 1.12 30 39.6 17 87 Intennediate Shell Plate A9153-2 Position 1.1 58 1.12 -20 65.0 34 79 Lower Shell Plate C99231 Position 1.1 51 1.12 10 57.1 34 101 Lower Shell Plate C9923 2 Position 1.1 51 1.12 10 57.1 34 101 Inter, & Lower Shell Position 2.1 36.5 0.838 -44 30.6 28 15 Longitudinal Weld Seams BC, BD and BA, BB (45' Azimuth)
Intermediate to Lower Shell Position 2.1 36.5 44 1.12 40.9 28 25 Circumferential Weld Seam AB Notes:
(a) Initial RTNDT values are measured values. l (b) ART = Initial RTNDT + ARTNDT + Margin (*F) i (c) ARTNDT = CF
- FF i
l Revision 1
4-12 TABLE 4-11 Calculation of the ART Values for the 3/4T Location @ 20 EFPY Material RG 1.99 R2 CF FF IRTu,/*' Margin ART *'
ARTuo/')
Method (*F)
Intermediate Shell Plate A91541 Position 2.1 35.4 0.860 30 30.4 17 77 Intermediate Shell Plate A9153 2 Position 1.1 58 0.860 -20 50.0 34 64 Lower Shell Plate C9923-1 Position 1.1 51 0.860 10 43.9 34 88
~
Lower Shell Plate C9923-2 Position 1.1 51 0.860 10 43.9 34 88 Inter & Lower Shell
' Pos tion 2.1 36.5 0.594 -44 21.6 21.6 -I Longitudinal Weld Seams BC, BD and BA, BB (45' Azimuth)
Intermediate to Lower Shell Pos tion 2.1 36.5 0.860 -44 3 a .4 28 15 Circumferential Weld Seam AB e
Notes:-
(a) Initial RTNDT values are measured values.
(b) ART = Initial RTNDT + ARTNDT + Margin ( F)
(c) ARTNDT = CF
- FF O
Revision 1
7 __
4-13 TABLE 4-12 I
Calculation of the ART Values for the 1/4T Location @ 32 EFPY Material RG 1.99 R2 CF FF IRTuor* M Margin A RT
ARTwor Method ('F)
Intermediate Shell Plate A9154-1 Position 2.1 35.4 1.24 30 43.9 17 91 intermediate Shell Plate A9153 2 Position 1.1 58 1.24 -20 71.9 34 86 Lower Shell Plate C9923-1 Position 1.1 51 1.24 10 63.2 34 107 Lower Shell Plate C9923-2 Position 1.1 51 1.24 10 63.2 34 107 Inter. & Lower Shell Position 2.1 36.5 0.970 -44 35.4 28 19 Longitudinal Weld Seams BC, BD and BA, BB (45* Azimuth)
Intermediate to Lower Shell Position 2.1 36.5 1.24 -44 45.3 28 29 Circumferential Weld Seam AB Notes: l (a) Initial RTNDT values are measured values.
(b) ART = Initial RTNDT + ARTNDT + Margin ( F)
(c) ARTNDT = CF
- FF 6
Revision 1
4-14 TABLE 4-13 Calculation of the ART Values for the 3/4T Location @ 32 EFPY Material RG 1.99 R2 CF FF IRTuo/*' ARTup/') Margin ART
- Method (*F)
Intermediate Shell Plate A9154-1 Position 2.1 35.4 0.986 30 34.9 17 82 Intermediate Shell Plate A9153-2 Position 1.1 58 0.986 -20 57.2 34 71 Lower Shell Plate C99231 Position 1.1 51 0.986 10 50.3 34 94 Lower Shell Plate C9923-2 Position 1.1 51 0.986 10 50.3 34 94 Inter. & Lower Shell Position 2.1 36.5 0.713 -44 26.0 26 8 Longitudinal Weld Seams BC, BD and BA, BB (45' Azimuth)
Intermediate to Lower Shell Position 2.1 36.5 0.986 -44 36.0 28 20 Circumferential Weld Seam AB Notes:
(a) Initial RTNDT values are measured values.
(b) ART = Initial RTNDT+ARTNDT + Margin ( F)
(c) ARTNDT = CF
- FF 9
Revision 1
4-15 The lower shell plates C9923-1 and C9923-2 are the limiting beltline materials for all heatup and cooldown curves to be generated. Contained in Table 4-14 is a summary of the limiting ARTS to be used in the generation of the V. C. Summer Unit I reactor vessel heatup and cooldown curves.
TABLE 4-14 Summary of the Limiting ART Values Used in the Generation of the V. C. Summer Unit 1 Heatup/Cooldown Curves EFPY 1/4T Limiting ART 3/4T Limiting ART 20 101'F 88 F s
32 107 F 94'F e
1 Revision 1
5-1 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Section 3 and 4 of this report.
Figures 5-1 and 5-3 present the heatup curves with margins of 10 F and 60 psig for possible instrumentation errors for heatup rates of 50 and 100*F/hr. The curves are applicable for 20 EFPY and 32 EFPY respectively, for the V. C. Summer Unit I reactor vessel. Additionally, Figures 5-2 and 5-4 present the cooldown curves without margins for possible instrumentation errors for cooldown rates of 0, 25,50 and 100*F/hr. These curves are also applicable for 20 EFPY and 32 EFPY, respectively, for the
' V. C. Summer Unit I 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 minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Appendix G to Section XI of the ASME Code [3] as follows:
1.5Ki. < Ku (12) where, K im is the stress intensity factor covered by membrane (pressure) stress, K Ic= 33.2 + 20.734 e (0.02 (T - RTNDT)],
T is the minimum permissible metal temperature, and RTNDT si the metal reference nil-ductility temperature The criticality limit curve specifies pressure-temperature limits for core operation to provide 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 minimum temperature required for the inservice hydrostatic test, and at least 40 F higher than the minimum permissible temperature in the corresponding pressure temperature curve for heatup and cooldown calculated as described in Section 3 of this report. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40 F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Revision 1
5-2 Figures 5-1 through 5-4 define all of the above limits for ensuring prevention of nonductile failure for the V. C. Summer Unit I 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 Revision 1
'5-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATES C9923-1,-2 LIMITING ART VALUES AT 20 EFPY: 1/4T,10l*F 3/4T, 88'F 2500 ,
^ ._ 1I50300457 i !4 I If f if i i i i i l
+
! i a I ;l 1i ;f is i i i i i ' e i e
! r i ! I fI i f
' f i '
tac 2 2 5 0 i e e i
I; L'Ar rust truir i i
' i 1 i i *
! I 9 I ! t ,
gn i i ' i i ll
- i ll iJ If f I e II ilmar! / II f: i
! l r
. ! e i i i i ! ! r . .I ! .i ' i 3 I .
! if 'il ' f- Je i a 2000 ' ' ' ' ' ' ' i' ' ' ' 'N ' EEATUP
. i v
i ! i i i ! i i i ! 1 J i'
Ii Ji i .' UP TO RATE
! i i !
i i !
- i i
i i i I If I r! 1i t t 3 50 F/Er. ' ! ! '
! 1 ! i t! I fi ! r! ! ! I i.
i i t/ t ti Ii i ti i i t I ! ', i C 1750~ UNACCEPTABLE
" ;/ I i,
li i Wu
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p OPERATION / I /. J l i r
UP TO 100 F/Er. .
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, i i i i
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5 .
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h/' / '\ \ l f
- I
\ \ CEIr. LIutT e i g . t i ! I i '
\ FOR i i ! ! !
y f i e i i i , 4 r > '
N 50 F/Er. r i i I e i i 750 4 . i i i
i i
x ,
i i i i
i i i i 1
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cc ! . i i ! i s emir. trust
' i , i 1 , , i 1 ,
m
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,i! '
i
! ! i i i ,
i s 500 '
J;' .
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ll!,' l': l','
U ' Bolt Up R I ' ' 'i! i *! i Temp. ,i il'N N i i i ' ,i i e
i ! I i e i i i j i , i ;j O 250 ! 'I i j i i i
I ! ' ' ! !'
CRITICALITY LIMIT BASED ON ! ' '
C) ! '
i
' INSERVICE BYDROSTA'IC TEST ' ' '
. t ii ! i ,'
' i 1! TENFERATURS (158 F FOR TRE t 1 1 i.
g ,,, , i is i SERVICE PERIOD UP 90 30.0 EFFY I ! i ! ! i
. i .
0 50 100 150 2$0 2$0 3$0 3$0 4$0 450 500 Moderator Temperature (Deg.F)
)
l FIGURE S-1 V. C. Summer Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 50 and 100 F/hr) Applicable to 20 EFPY (Without Margins for Instrumentation Errors) 1 Revision 1 !
I
5-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATES C9923-1,-2 LIMITING ART VALUES AT 20 EFPY: 1/4T,10l*F 1 3/4T,88'F 2500 -
i .
i, , ,
- ssessees, i i i! ! ! ,
! i i. >
.i i i ,i i , . , ,.i! . , i
^ t i I '
i I ! I! i ' ! t i ! ! ! i ! ! i i3 ' ' i '
fll um 2250
~
l l! ,
l,' l', l, ; ;l. ;
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~
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f i i F/Er. j i ii e i
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e
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I ce 250 Temp. , . l,,i ;
. O
. i, l'i' llli i i i,.!
0-0 50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F)
FIGURE 5-2 V. C. Summer Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100'F/hr) Applicable to 20 EFPY (Without Margins for Instrumentation Errors)
Revision 1 1
1
5-5 TABLES-1 V. C. Summer Unit i Heatup Data at 20 EFPY (Without Margins for Instrumentation Errors) ,
Configuration # ll5030 W 7 50 'F/hr Critical. Limit 100 F/hr Critical. Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Pnss.
('F) (psig) ('F) (psig) (*F) (psig) ( F) (psig) ('F) (prig) 60 0 158 0 60 0 158 0 140 2000 60 809 158 828 60 809 158 828 158 2485 65 828 158 921 65 828 158 906 70 848 158 930 70 848 158 899 75 870 158 945 75 870 158 897 80 895 158 967 80 895 158 900 35 921 158 995 85 897 158 908 90 930 158 1029 90 897 158 920 95 945 158 1068 95 897 158 937 100 967 160 1113 100 900 160 959 105 995 165 1163 105 908 165 985 110 1029 170 1220 110 920 170 1016 115 1068 175 1284 115 937 175 1053 120 1113 180 1355 120 959 180 1094 I?5 1163 185 1434 125 985 185 1142 130 1220 190 1521 130 1016 190 1195 135 1284 195 1618 135 1053 195 1255 140 1355 200 1725 140 1094 200 1322 145 1434 205 1843 145 1142 205 1397 150 1521 210 1974 150 1195 210 1481 155 1618 215 2119 155 1255 215 1574 160 1725 220 2279 >50 1322 220 167o 165 1843 225 2455 165 1397 225 1791 170 1974 170 1481 230 1917
~
175 2119 175 1574 2056 235 _
180 2279 180 1676 240 2211 185 2455 185 1791 245 2381 190 1917 195 2056 200 2211 205 2381 I
Revision 1
5-6 TABLE 5-2 V. C. Summer Unit 1 Cooldown Data at 20 EFPY (Without Margins for Instrumentation Errors)
Configuration # 1150309457 Steady State 25 *F/hr 50 'F/hr 100 F/hr Temp. Press. Temp. Press. Temp. Press. Temp. Press.
( F) (psig) ('F) (psig) ( F) (psig) ( F) (psig) 60 0 60 0 60 0 60 0 60 809 60 774 60 740 60 675 65 828 65 794 65 762 65 702 70 848 70 817 70 787 70 731 75 870 75 841 75 814 75 764 80 895 80 868 80 843 80 801 4 85 922 85 898 85 876 85 841 l 90 953 90 932 90 913 90 886
{
95 986 95 968 95 953 95 935 100 1023 100 1009 100 998 100 990 105 1064- 105 1054 105 1048 105 1051 ,
110 1109 110 1104 110 1103 115 1159 115 1159 ,
120 1214 1 125 1275 130 1343 135 1417 l 140 1499 145 1590 150 1691 155 1802 160 1925 165 2060 170 2210 175 2376 Revi* n1 L.
1 5-7 MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATES C9923-1,-2 LIMITING ART VALUES AT 32 EFPY: 1/4T,107 F 3/4T,94 F 2500 ,
A
- sei T.. .. i i < ! i; i, i > i i i i
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i li ii !I i l i i i ! i 1
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LEAK TEST LIMIT '
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t
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(
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MEATUP RATE i UNACCEPTABLE il i f i
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li I i
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7 i UP TO 100 F/Er. ~ET 1
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iia i i t * . . i i i ! . 7 #
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" 750 ' ' ' ' ' ' '
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FOR 100 F/Er. !
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f I i , ,
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Temp. i e i t ! .t :
x ;
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i ; i , ,
O 250 ' ' '
i i i , i i CJ
! ' I I CRITICALITY LIJIT BASED ON ! ' ' '
' I INSERVICE BYD305fAUIC TEST ' !
, i
,'i !i . f.'_.TEMPERATURE BERTICE PERIOD UP P0 35.0 BFFT (194 F I FOR TER i I
i * .i 0 i i
. i ', . ,
0 50 100 150 2b0 250 300 3$0 4d0 450 500 Moderator Temperature (Deg.F)
FIGURE 5-3 V. C. Summer Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 50 and 100 F/hr) Applicable to 32 EFPY (Without Margins for anstrumentation Errors)
Revision 1
l 5-8 MATERIAL PROPERTY BASIS '
1 LIMITING MATERIAL: LOWER SHELL PLATES C9923 1,-2 LIMITING ART VALUES AT 32 EFPY: 1/4T,107 F 1 3/4T,94'F
- 2500 ,
I i ,
i i ii
,,. . . , , , , .., j 4, i i , . ii, . i i .
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I
. i i i i
Temp.
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,, , , i , , i , l'! l : , l,,'
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! I
' ' ' ' t 1 i ( ! i I i 0 ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '
~
0 50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F)
FIGURE 5-4 V. C. Summer Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100 F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors) '
Revision 1
5-9 TABLE 5-3 V. C. Summer Unit 1 Heatup Data at 32 EFPY (Without Margins for Instrumentation Errors) i Configuration # 2010760105 50 *F/hr Critical. Limit 100 F/hr Critical. limit Leak Test Limit i Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.
( F) (psig) ('F) (psig) ( F) (psig) ( F) (psig) ( F) (psig) 60 0 164 0 60 0 164 0 146 2000 60 789 164 806 60 789 164 806 164 2485 65 806 164 887 65 806 164 873 70 824 164 893 70 824 164 864 l 75 844 164 906 75 844 164 861 80 866 164 925 80 ' 861 164 863 85 887 164 949 85 861 164 868 90 893 164 978 90 861 164 878 95 906 164 1013 95 861 164 893 100 925 164 1052 100 863 164 911 105 949 165 1097 105 868 165 934 110 978 170 1147 110 878 170 961 115 1013 175 1204 115 893 175 992 120 1052 180 1267 120 911 180 1029 125 1097 185 1336 125 934 185 1070 130 1147 190 1414 130 961 190 1117 135 1204 195 1500 135 992 195 1170 140 1267 200 1595 140 1029 200 1230 145 1336 205 1700 145 1070 205 1296 150 1414 210 1816 150 1117 210 1370 155 1500 215 1944 155 1170 215 1452 160 1595 220 2085 160 1230 220 1543 165 1700 225 2242 165 1296 225 1644 170 1816 230 2415 170 1370 230 1755 175 1944 175 1452 235 1879 180 2085 180 1543 240 2016 185 2242 185 1644 245 2167 190 2415 190 1755 250 2334 195 1879 200 2016 205 2167 210 2334 Revision 1
i 5-10 1
1 TABLE 5-4 V. C. Summer Unit 1 Cooldown Data at 32 EFPY (Without Margins for Instrumentation Errors)
Configuration # 2010760105 I Steady State 25 'F/hr 50 *F/hr 100 F/hr Temp. Press. Temp. Press. Temp. Press. Temp. Press. l
(*F) (psig) (*F) (psig) (*F) (psig) ('F) (psig) l 60 0 60 0 60 0 60 0 l , 60 789 60 753 60 717 60 646 65 806 65 770 65 736 65 670 70 824 70 790 70 757 70 696.
75 844 75 812 75 781 75 725
_ 80 866 80 836 80 808 80 757 ,
j 85 890 85 862 85 837 85 792 90 917 90 892 90 869 90 832 l 95 946 95 925 95 905 95 876 i 100 979 100 961 100 945 100 925 I 105 1015 105 1000 105 989 105 979 110 1056 110 1044 110 1037 110 1039 115 1100 115 1093 115 1091 l 120 1149 120 1147
! 125 1203 l 130 1263 135 1329 l 140 1402 145 1482 150 1571 155 1670 160 1779 165 1899 170 2032 175 2179 180 2341 Revision 1
r 6-1 6 REFERENCES 1 Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S.
Nuclear Regulatory Commission, May,1988.
2 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", Federal Register, Volume 60, No. 243, dated December 19,1995.
3 1996 ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Appendix G, " Fracture Toughness Criteria for Protection Against Failure", December 1996.
4 ASME Code Case N-640," Alternative Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1", February.
5 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.
6 1989 Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331,
" Material for Vessels".
7 WCAP-12867, " Analysis of Capsule X from the South Carolina Electric & Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program", J. M. Chicots, et al., March !
1991.
- 8. WCAP-15101, Arulysis of Capsule W from the South Carolina Electric & Gas V. C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program", T. J. Laubham, et al., September 1998.
l 9 CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.1, developed by ATI Consulting, March 1996.
10 WCAP-15102, Rev. O, "V. C. Summer Unit 1 Heatup and Cooldown Limit Curves for Normal Operation", S. Spragg, et. al., September 1998.
1 Revision 1
,