ML20210T571
| ML20210T571 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/30/1986 |
| From: | Congedo T, Gong H, Palusamy S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20210T530 | List: |
| References | |
| TAC-60075, WCAP-10934, WCAP-10934-R02, WCAP-10934-R2, NUDOCS 8610090078 | |
| Download: ML20210T571 (32) | |
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ATTACHMENT 2 WCAP-10934 Rev. 2 , WESTINGHOUSE CLASS 3 CUSTOMER DESIGNATED DISTRIBUTION HEATUP AND C00LDOWN LIMIT CURVES FOR THE ALABAMA POWER COMPANY JOSEPH M. FARLEY UNIT 1 REACTOR VESSEL H. Gong T. V. Congedo S. E. Yanichko T. R. Mager June 1986 Approved: , 4 ~e/ b M S. S. Pa ysamy, Manager Struct #al Materials Engineering Prepared by Westinghouse for the Alabama Power Company.
Work Performed Under Shop Order AIVJ-139 Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.
Westinghouse Electric Corporation Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230 8610090078 860929 PDR ADOCK 05000348 P PDR 91190:10/070786
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- 1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature). The most limiting RT NDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART NDT. RT NDT is designated as the higher of either the drop weight nil-ductility transition temperature (TNDT) r the temperature at which the material exhibits at least 50 ft ib of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.
RT NDT increases as the material is exposed to fast-neutron radiation. Thus, to find the most limiting RT at any time period in the reactor life, NDT ART due to the radiation exposure associated with that time period must NDT be added to the original unirradiated RT NDT. The extent of the shift in RT NDT is enhanced by certain chemical elements (such as copper, nickel and phosphorus) present in reactor vessel steels. Westinghouse, other NSSS vendors, the U.S. Nuclear Regulatory Commission and others have developed trend curves for predicting adjustment of RT NDT as a function of fluence and copper, nickel and/or phosphorus content. The Nuclear Regulatory Comission (NRC) trend curve is published in Regulatory Guide 1.99 (Effects of Residual Elements on Predicting Radiation Damage to Reactor Vessel Materials)III.
Regulatory Guide 1.99 was originally published in July 1975 with a Revision 1 being issued in April 1977. Currently, a Revision 2 to Regulatory Guide 1.99 is under consideration within the NRC. The chemistry factor, "CF", 'F, a function of copper and nickel content identified in Regulatory Guide 1.99, Revision 2 is given in Table I for welds and Table 11 for base metal (plates and forgings). Interpolation is permitted. The value, "f", given in Figure 1 is the calculated value of the neutron fluence at the location of interest (inner surface, 1/4T, or 3/41) in the vessel at the location of the postulated defect, n/cm2 (E > MeV) divided by 10I9 . The fluence factor is determined from Figure 1.
9119Q: 10/061786 Given the copper and nickel contents of the most limiting material, the radiation-induced ART NDT can be estimated from Tables I and II and Figure
- 1. The maximum fast-neutron fluence (E > 1 MeV) at the inner surface,1/4T (wall thickness) and 3/4T (wall thickness) vessel locations is given as a function of full-power service in Figure 2 and 3 for.the vessel core region girth weld and longitudinal welds', respectively, and plates shown in Figure 4. The data for all other ferritic materials in the reactor coolant pressure boundary are examined to ensure that no other component will be limiting with respect to RT NDT*
- 2. FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture-toughness properties of the Farley Unit I reactor
, vessel materials are presented in Table III. A review of submerged arc welding practice at the time of vessel fabrication by Combustion Engineering showed that B-4 type weld wire was used and the as deposited nickel content of the welds resulted in low nickel (~0.20% Ni). The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan (2) . The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the Farley Unit i Vessel Material Surveillance Program.
- 3. FLUENCE CALCULATIONS For the purpose of revising heatup and cooldown curves for Farley Unit 1, it is necessary to know vessel fast fluence (+ (E > 1 MeV)) at the azimuthal peak location which has limiting embrittlement characteristics. This peak location is at O' for plates and girth welds and 45' for longitudinal weld seams shown in Figure 4. The calculations performed for this purpose consist of adjoint analyses, relating the fast flux (+ (E > 1 MeV)) at the vessel IR to the power distributions in the reactor core. The adjoint (importance)
- functions used, when combined with cycle specific core power distributions, yield the plant specific exposure data for each operating fuel cycle.
The adjoint function was generated using the 00T discrete ordinates code (3) and the SAILOR cross-section library I4) . The SAILOR library is a 47 group, i ENDF-B/IV based data set produced specifically for light water reactor I
applications. In generating the adjoint function, anisotropic scattering was 91190:10/061986 .. --~
treated with a P expansion of the cross-sections. The adjoint source 3
location was chosen along the inner diameter of the pressure vessel. This calculation was run in R, O geometry to provide a power distribution importance function for the exposure parameter of interest ($ (E > 1 Mey)).
Having the adjoint importance function and appropriate core power distributions, the response of interest is calculated as RR,0 " IR I0 1(R,0) F(R,0) R dR de where:
R - Response of interest ($ (E > 1.0 MeV), dPa, etc.) at radius R,0 R and azimuthal angle 0.
I(R,0) - Adjoint impprtance function at radius R and azimuthal angle 0.
F(R,0) - Full power fission density at radius R and azimuthal angle G.
It should be noted that as written in the above equation, the importance function 1(R,0) represents an integral over the fission distribution so that the response of interest can be related directly to the spatial distribution of fission density within the reactor core.
Core power distributions for Farley Unit I were taken from the following Westinghouse fuel cycle design reports for each operating cycle to date:
Fuel Cycle Report 1 WCAP-8515 and Ref. 5 2 Ref. 5 and WCAP-9761 3 WCAP-9761 and WCAP-10036 4A WCAP-10036 and WCAP-10308 5 WCAP-10308 and WCAP-10525 6 WCAP-10525 and WCAP-10795 7 WCAP-10795 and Ref. 6 9119Q: 10/061786 _. ._ . _ .-_ _ - .- - . -_ _ _ - _
Of these, Cycles 1 through 4A utilized out-in fuel loading patterns, and Cycles 5, 6 and 7 implemented low leakage fuel loading patterns.
The power distributions employed represent cycle averaged relative assembly powers. Therefore, the adjoint results are in terms of fuel cycle averaged neutron flux, which when multiplied by the fuel cycle length yields the incremental fast neutron fluence. Fast fluences at 1/4T and 3/4T are obtained from those at vessel IR through fast flux ratios obtained from the 00T transport analysis performed in support of WCAP-10474, " Analysis of Capsule U from the Alabama Power Company, Joseph M. Farley Unit 1 Reactor Vessel
, Radiation Surveillance Program". As a result, the following neutron fluences for E > 1.0 MeV were calculated:
Cumulative fluence (E > 1 MeV) at O' Lifetime ,
(n/cm )
EFPY Vessel IR Vessel 1/4T Vessel 3/4T 1.943 x 10 18 2.674 x 10 I 18 1.05 1.151 x 10 1.81 3.526 x 10 18 2.088 x 10 18 4.852 x 10 lI 2.19 4.231 x 10 18 2.505 x 10 18 5.822 x 10 II 2.98 5.738 x 10 18 3.398 x 10 18 7.896 x 10 lI 18 18 3.81 7.041 x 10 4.169 x 10 9.688 x 10 II 4.72 8.296 x 10 18 4.913 x 10 18 1.142 x 10 18 4.79 8.406 x 10 I0 4.978 x 10 18 1.157 x 10 18 32.00 5.037 x 10 I9 2.983 x 10 I9 6.932 x 10 18 Projection of fluence at 32.0 EFPY was made assuming a power distribution unchanged from that used in Cycle 7 (i.e., that which generated the fluence estimate for 4.79 EFPY).
9119Q:lD/061786 4-
- 4. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg , for the combined thermal and pressure stresses at any time during heatup and cooldown cannot be greater than the reference stress intensity factor, KIR, f r the metal temperature at that time. K IR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME CodeI I. The K IR curve is given by the equation:
KIR = 26.78 + 1.223 exp (0.0145 (T-RTNDT + 160)) (1) where K yp is the reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT Thus, the governing equation of the heatup-cooldown analysis is NDT.
defined in Appendix G to the'ASME CodeII as follows:
CKgg + kit # K IR (2) where:
K yg is the stress intensity factor caused by membrane (pressure) stress K
yg is the stress intensity factor caused by the thermal gradients K
gp is a function of temperature relative to the RT NDT f 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.
91190: 10/061786 I
At any time during the heatup or cooldown transient, K IR is deternined by the metal temperature at the tip of the postulated flaw, the appropriate value of RTNDT, and the' reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, kit, f r 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 Code reference flaw 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. Frora these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on 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 10. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4T IR location for finite cooldown rates than for steady-state operation.
Furthermore, if conditions exist such that the increase in K exceeds IR Kh , 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.
9119Q: 10/061786 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 vessel wall. The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K g for the 1/4T crack during heatup is lower than the K yp 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 such that the effects of compressive thermal stresses and lower K IR s 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 consid,ered. Therefore, 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 pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw 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 thus 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 as follows: A composite curve is constructed 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 91190: 10/061786 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. Then, composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves in Figures 5 through 8. In addition, heatup and cooldown curves without instrument errors are presented in Figures 9 through 12.
i
) The Farley Unit 2 fracture analysis results from Reference 8 are applicable to
]
Farley Unit 1 since the pertinent parameters are identical for both plants.
As a result, the 16 EFPY and 32 EFPY cooldown curves with and without instrument errors are impacted by the 10CFR50 rule as shown by Figures 6, 8, l
10, and 12; also, the heatup curves without instrument errors for 16 and 32
' EFPY and with instrument error for 16 EFPY are impacted by the 10CFR50 rule as shown by Figures 5, 9 and 11. However, the heatup curve with instrument errors for 32'EFPY shown by Figure 7 is not impacted by the 10CFR50 rule.
Since there are many conservatisms (safety factor of 2 on pressure, K IR toughness and 1/4T flaw) built into the ASME Appendix G analysis method I) ,
Appendix G does not require that instrument error margins be included in the analysis. Therefore, plant operation can be based on heatup and cooldown curves without instrument errors.
An evaluation has been performed to determine the acceptability of the Overpressure Mitigation System (OMS) presently in Farley Unit 1 (Technical Specification 3/4.4.10.3) with respect to the 16 EFPY heatup and cooldown curves shown in Figures 9 and 10 respectively. For the purpose of the evaluation it was assumed that the RHR relief valve lifts at 495 psig which includes 10% accumulation. The heatup curve in Figure 9 does not fall below ,
495 psig at any temperature. A comparison of cooldown curves in Figure 10 shows that in the low temperature range (<130*F) cooldown rates of 20*F/hr and lower fall well above 495 psig. Although the cooldown curves for rates of
, 40*F/hr and above do f all below 495 psig, it is not expected that the Appendix -
G curves will be violated during an actuation of the OMS since cooldown rates greater than or equal to 40*F/hr are highly unlikely at low temperature 91190:10/070786 .. .
~ _ _ _ . _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . - _ . _ _ _ . _ .
conditions. Therefore, the Appendix G curves as illustrated in Figures 9 and 10 will not be violated as the result of an actuation of the OMS.
- 5. HEATUP AND C00LDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed previously. The derivation of the limit curves is presented in the NRC Regulatory Standard Review Plan (2) ,
Transition temperature shifts occurring in the pressure vessel materials due to radiation exposure have been obtained directly from the reactor pressure vessel surveillance program.
Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown on the heatup and cooldown curves. The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line, shown in Figures 5, 7, 9 and 11. This is in addition to other criteria which must be met before the reactor is made critical.
The leak test limit curve shown in Figures 5, 7, 9 and 11 represent minimum temperature requirements at the leak test pressure specified by applicable codes (2,7) ,
- 6. AVAILABLE SURVEILLANCE CAPSULE DA1A AND ADJUSTED REFERENCE 1EMPERATURE Credible surveillance data is available for lower plate B6919-1
,10]
however, data for the weld metal is not considered credible since it was not fabricated with the same heat of weld wire and lot of flux as used in the vessel beltline region. The following surveillance data was used to determine a chemistry factor (CF) for plate B6919-1 using USNRC Regulatory Guide 1.99 3
Rev. 2.
91190:10/061986 ._ _ . _ _ _ _ _ _ - - _ _ . _ _ _ . . . _ _ _ _ . _ .___ _.__ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _
From Regulatory Guide 1.99 Rev. 2 the adjusted reference temperature (ART) for each material in the beltline is given by the following expression:
ART = Initial RTNDT + ARTNDT + argin (3)
Initial RT NDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. If measured values of Initial RT NDT 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.
ART is the mean value of the adjustment in reference temperature caused ND1 by irradiation and should be calculated as follows:
ART surf ace = (CF]f(0.28-0.10 log f) (4)
NDT To calculate ART NDT at any depth (e.g., at 1/41 or 3/41), the following attenuation formula was used:
NDT where x (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface.
CF (*F) is the chemistry factor, a function of copper and nickel content. CF is given in Table I for welds and in Table II for base metal (plates and forgings). Linear interpolation is permitted. In Tables I and II
" 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 for weld samples made with the weld wire heat number that matches the critical vessel weld.
91190: 10/061986 __ _. - - _ . - - . - _ _ _ .-
I Surveillance Base Material Data (Plate B6919-1)
ART Fluence Factor ART x FF 2 NDT gpp)
Fluence (n/cm ) *F NDT 5.83 x 10 18 55 0.849 46.695 1.65 x 10 I8 90 1.138 102.420 18 72.165 5.83 x 10 85 0.849 I 119.49 1.65 x 10 ' 105 1.138 340.77 Sum of the squares of the fluence factor = 4.0317.
- 0 7 = 84.54 Then Chemistry Factor (CF) = ,
9 7 I (FF)p i
Beltline Base Material Evaluation 32 EFPY ART NDT I I Initial Plate No. RTNDT(*F) Cu Ni CF Surface 1/4 T* 3/4 T*
B6903-2 0 .13 .60 91 127.8 112.0 86.0 B6903-3 10 .12 .56 82.2 115.4 101.1 77.7 86919-1 15 .14 .55 84.5 118.7 104.0** 79.9**
B6919-2 5 .14 .56 98.2 137.9 120.9 92.8
- aRTNDT = [ARTNDT urface]e
- Base on surveillance data Margin =2)o2 , ,2, 2
2 \ 0 + 17 = 34*F ART = Initial RTNDT + ARTNDT + argin 91190: 10/061986 . - . _ _ _ . .__ _ _ _ _ - , _ _ - , _ _ _ _ _ _ _ . _ _-._-._ .._ _ _ -... . .._.,_._ _
Initial Plate No. RTNDT(*F) Marain (*F) 1/4 T* 3/4 T*
B6903-2 0 34 146.0 120.0 B6903-3 10 34 145.1 121.7 B6919-1 15 34 136.0* 111.9*
B6919-2 5 34 159.9 1 31.8
- Based on surveillance data Weld Metal Evaluation 32 EFPY Initial Weld No. RTNDT(*F) Cu Ni CF Surface
- 1/4 T** 3/4 T**
19-894A&8 -56 .25 .21 127.1 142.1 124.5 95.7 11-894 -56 .22 .20 112.0 157.2 137.8 105.8 20-894A&B -56 .17 .20 92.0 102.8 90.1 69.2 I* ~
- 9 }
- ARTNDT [CF]f
Initial Weld No. RTNDT(*F) Marain (*F) 1/4 T 3/4j e 19-894A&B -56 65.5 134.0 105.2 11-894 -56 65.5 147.3 115.3 20-894A&8 -56 65.5 99.6 78.7 91190: 10/061986 - _ - _ _ _ _ _ _ _ - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ .b
Data for 16 EFPY was also generated as above. Based on the above analysis, plate B6919-2 is considered the limiting material and was used to develop heatup and cooldown curves shown in Figures 5 through 12.
f
- 7. SURVEILLANCE CAPSULE REMOVAL SCHEDULE The surveillance capsule withdrawal schedule for Unit 1 (Table IV) should remain the same as identified in the Technical Specifications and WCAP-10474 I9) . The dosimetry analysis of the third capsule to be removed af ter 6 EFPY should be used to re-evaluate the withdrawal schedule for the remaining capsules.
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j-9119Q:10/070786 _ _ _ _ - . _ _ _ _ _ _ . - _ _ _ _ _ _ . - _ _ _ _ - ~ _ _ . _ _ . - _ _ _ . _ . _ . _ _ _ . . _ - _ . _
s REFERENCES (1) Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on ,
Predicted Radiation Damagi to Reactor Vessel Materials," U.S. Nuclear '
Regulatory Commission, April 1977.
(2) " Fracture Toughness Requireraents," Branch Technical Position - MTEB No.
5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for N clear Power Plants, LWR Edition, NUREG-0800, 1981.
(3) Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation Vol. 5 - Two Dimensional, Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.
(4) " Sailor RSIC Data Library Collection DLC-76," Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P , Cross Section Library foi Light Water 3
Reactors, Radiation Shielding Information Center, Oak Ridge National Laboratory. ,
(5) Radcliffe, R. and Holmes, R., " Revised Nuclear Design Dati for Cycle 2 of FarleyUnit1,"WestinghouseNuclearFuelsDivikion,N05-79-096, April 25, 1979. <;
(6) Erwin, R., "New Power Distributions for Cycle 7," J. M. Farley Unit 1 Project File, Westinghouse Nuclear fuels Division, August 19, 1985.
(7) ASME Boiler and Pressure Vessel Code,Section III, Division 1 - ~
j Appendices, " Rules for Construction of Nuclear Vessels," Appendix.G,
" Protection Against Nonductile Failure," pp. 559-564, 1983 Edition, American Society of Mechanical Engineers, New York, 1983.
(8) Miller, J. C., " Response to NRC Comments on Farley Unit 2 " ALA-85-706, July 31, 1985.
(9) Boggs, R. S., Yanichko, S. E., Cheney, C. A. and Kaiser, W. T. " Analysis of Capsule U from the Alabama Power Company Joseph M. Far ley' Unit 1 I
Reactor Vessel Radiation Surveillance: Program," WCAP-10474. February 1984.
9119Q:lD/061786 ?
(10) Yanichko, S. E., Anderson, S. L., and Kaiser, W. T., " Analysis of Capsule Y from the Alabama Power Company Farley Unit No. 1 Reactor Vessel Radiation Surveillance Program," WCAP-9717, June, 1980.
(11) Regulatory Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials" (Proposed Draft),
U.S. Nuclear Regulatory Commission, June 1984.
9119Q: 10/061786 TABLE I CHEMISTRY FACTOR FOR WELOS, 'F Copper, Nickel, Wt. %
Wt. % 0 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02 21 26 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 24 43 54 54 54 54 54
- 0.05 26 49 67 68 68 68 68 0.06
~
29 52 77 82 82 82 82
'O.07 32 55 85 95 95 95 95 0.08 36 58 90 106 108 108 108 0.09 40 61 94 115 122 122 122 0.10 44 65 97 122 133 135 135 0.11 49 68 101 130 144 148 148 0.12 52 72 103 135 153 161 161 0.13 58 76 106 139 162 172 176 0.14 61 79 109 142 168 182 188 0.15 66 84 112 146 175 191 200 O.16 70 88 115 149 178 199 211 0.17 75 92 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191 220 238 b.20 88 104 129 160 194 223 245 0.21 92 108 133 164 197 229 252 0.22 97 112 137 167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 239 268 0.25 110 126 148 176 209 243 272 0.26 113 130 151 180 212 246 276 O.27 119 134 155 184 216 249 280 0.28 122 138 160 187 218 251 284 0.29 128 142 164 191 222 254 287 0.30 131 146 167 194 225 257 290 0.31 136 151 172 198 228 260 293 0.32 140 155 175 202 231 263 296 0.33 144 160 180 205 234 266 299 0.34 149 164 184 209 238 269 302 0.35 153 168 187 212 241 272 305 0.36 158 172 191 216 245 275 308 0.37 162 177 196 220 248 278 311 0.38 166 182 200 223 250 281 314 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 91190: 10/061786 ,
TABLE II CHEMISTRY FACTOR FOR BASE METAL, 'F Copper, Nickel, Wt. %
Wt. % 0 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20 0.02 20 23 20 20 20 20 20 0.03 20 20 20 20 20 20 20 0.04 22 26 26 26 26 26 26 0.05 25 31 31 31 31 31 31 0.06 28 27 37 37 37 37 37 0.07 31 43 44 44 44 44 44 0.08 34 48 51 51 51 51 51 O.09 37 53 58 58 58 58 58 0.10 41 58 65 65 67 67 67 0.11 45 62 72 74 77 77 77 0.12 49 67 79 83 86 86 86 0.13 53 71 85 91 96 96 96 0.14 57 75 91 100 105 106 106 0.15 61 80 99 110 115 117 117 0.16 65 84 104 118 123 125 125 0.17 69 88 110 127 132 135 135 0.18 73 92 115 134 141 144 144 0.19 78 97 120 142 150 154 154 0.20 82 102 125 149 159 164 165 0.21 86 107 129 155 167 172 174 0.22 91 112 134 161 176 181 184 0.23 95 117 138 167 184 190 194 0.24 100 121 143 172 191 199 204 0.25 104 126 148 176 199 208 214 0.26 109 130 151 180 205 216 221 0.27 114 134 155 184 211 225 230 0.28 119 138 160 187 216 233 239 0.29 124 142 164 191 221 241 248 0.30 129 146 167 194 225 249 257 0.31 134 151 172 198 228 255 266 0.32 139 155 175 202 231 260 274 0.33 144 160 180 205 234 264 282 0.34 149 164 184 209 238 268 290 0.35 153 168 187 212 241 272 298 0.36 158 173 191 216 245 275 303 0.37 162 177 196 220 248 278 308 0.38 166 182 200 223 250 281 313 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 9119Q: 10/061786 TABLE III FARLEY UNIT 1 REACTOR VESSEL TOUGHNESS PROPERTIES Material Cu P Ni T NDT RT NDT Upper Shell Energy Component Code No. Type (%) (%) (%) (*F) (*F) MWO(C) NMWD(d)
Closure head dome B6901 A533.B.C1.1 0.16 0.009 0.50 -30 -20[a] 140 -
Closure head segment B6902-1 A533,B.C1.1 0.17 0.007 0.52 -20 -20[a] 138 -
Closure head flange B6915-1 A508, C1.2 0.10 0.012 0.64 60[a] 60[a] 75[a] _
Vessel flange 86913-1 A508, C1.2 0.17 0.011 0.69 '
60[a] 60[a] 106[a] _
Inlet nozzle B6917-1 A508, C1.2 -
0.010 0.83 60[a] 60[a] -
110 Inlet nozzle B6917-2 A508, C1.2 -
0.008 0.80 60[a] 60[a] -
80 Inlet nozzle B6917-3 A508, C1.2 -
0.008 0.87 60[a] 60[a] -
98 Outlet nozzle B6916-1 A508, C1.2 -
0.007 0.77 60[a] 60[a] -
96.5 Outlet nozzle B6916-2 A508, C1.2 -
0.011 0.78 60[a] 60[a] -
97.5 Outlet nozzle B6916-3 A508, C1.2 -
0.009 0.78 60[a] 60[a] -
100
'? Nazzle shell B6914-1 A508, C1.2 -
0.010 0.68 30 30[a] 148 -
Inter. shell B6903-2 A533,B,C1.1 0.13 0.011 0.60 0 0 1 51 .5 97 Inter. shell B6903-3 A533,B,C1.1 0.12 0.014 0.56 10 10 134.5 100 Lower shell 86919-1 A533,B,C1.1 0.14 0.015 0.55 -20 15 133 90.5 Lower shell B6919-2 A533,B,C1.1 0.14 0.015 0.56 -10 5 134 97 Bsttom head ring B6912-1 A508, C1.2 -
0.010 0.72 10 10[a] 163.5 -
Bottom head segment B6906-1 A533,B,C1.1 0.15 0.011 0.52 -30 -30[a] 147 _
Bottom head dome B6907-1 A533,B,C1.1 0.17 0.014 0.60 -30 -30[a] 143.5 -
Inter. shell long. M1.33 Sub Arc Weld 0.25 0.017 0.21 0[a] 0[a] _ _
i weld seam Inter. to lower G1.18 Sub Arc Weld 0.22 0.011 <0.20[b] o[a] o[a] _ _
shell weld seams Lower shell long. G1.08 Sub Arc Weld 0.17 0.022 <0.20[b] 0[a] o[a] _ _
weld seams i
l [a] Estimate per NUREG-0800 "USNRC Standard Review Plan" Branch Technical Position MTEB 5-2.
[b] Estimated (low nickel weld wire used in fabricating vessel weld seams).
[c] Major working direction.
[d] Normal to major working direction i
. . . . . , ~ , . ,
l TABLE IV i
SURVEILLANCE CAPSULE REMOVAL SCHEDULE Lead Estimated Fluence 2 I9 Capsule Factor Removal Time [a] n/cm x 10 Y 3.12 Removed (1.13) .583 (Actual)
U 3.12 Removed (3.02) 1.65 (Actual)
X 3.12 6 3.05[b]
W 2.70 12 5.28[c]
V 2.70 21 9.28 Z 2.70 Standby -
[a] Effective full power years from plant startup
[b] Approximates vessel end of life 1/4 thickness wall location fluence
[c] Approximates vessel end of life inner wall location fluence l
i l
l l
9119Q: 10/061786 .)g, 1
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i i .
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. 8-:---= -
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. 3 4 5 ( ' S9 3 4 5 . 7891 4 5 789 If_ ' 8 10Je 10:e ig 32 i Fluence, n/cm8 lE > 1 MeV) 4 i
i 1
FIGURE 1 .
! FLUENCE FACTOR FOR USE IN THE EXPRESSION FOR ART 4 NDT I
h l
- . i
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l i
i
T 2
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- 6 .. .a . + .. .l 4 --
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ts
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7" . ....
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= = ,___... Q __. _. ._..;.;
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. _.s , _. .p:- : =.m l% . p. .9 g _., _m , i e . ,.csi.. 1 . :._ _ - ,-. . e 5 : .. ._. + w 9
g N_*. . .'. I i f.i *.- FE4 i [ :. ; i '.*_ : [ .i ! .1 - , 1 f4 b:.: : -'-
' ( ; .'{ & s. N : ..- L !. : 1. :i1:6
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. :... :.: . . . :_ . _ = ,. e. : : . . _..:. r. _ __..._.,...-:
. . . . . . . . _ . _ . . ._.. .. .. _ . . . ... _.. .. .. _. .y. . s . . :- u__:..__-=...
7'
. NEUTRON .FLllkNCE- MH- Ne
@ASi-MEUNCfION.ONFULL R0WERL M s ' -W/:= .FI6uREi.2giEAST
._... .. _. __ __ .. .__.__ _. .. __.. . mu u.
,i._._. .. __ __. __ _ . . . .. . . . . -
\
- - . . _ . m_r__. u.. . _ _.
- 5. .
.u .., m f. s. u . , ; ; , e.; -,.; = m t . 1 . .. ..a . .; . . : 4; m. ,.,._...,m. . _. .
__.: : -: g.:.. r,.._._. _.
, 1 . . .. , = . ,. . . _ , . . . ... . . .: __ _ .-...i.... .. . . .
- 5d@ :p i ti-i E%L F::-W:-Vi I 3 I . ' 5 .i-t- T : :-T i i-L4 I? H '.. ..4 M_ _ __. . . _"_f l
'..l.... . ..___ _. ._. .____ __. __ .._. _. ... _. __ .. _...._..c.__.._ . . . .K_ _._
. . __ 1_2
_.....r.-.. ....__. .,.. _.... .. . _._.. _.._.... .. . ..
3.
. _ _ . .. ... .....t,.... .._c....___ .__
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_ r._._ . .____
2.
.. 5.____ -
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. . . _ . ....tr i
l 17 -
10 ,-
0 5 10 15 20 25 30 35 EFPY l
I
2 10.0 - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . . . _ . _ _ . _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _
= ----;
9... -_c------- x -- - - - - -
8...
7... _ _ _
6...
5...
4...
3,,, _ _ _
1 I
2...
m SURFAEE i 4 , , f
' t g a f , g I 6 i l y . i I i ! l # I i i g ' ' ' ' # 8 10 .. :
J4tE=-
--__ _r 9.. - - - --- -
B.. _-
7..
m n
6 6.-
~.-'-' '
%e 5 .. - -
w w a__
,-- __s o _- -
z ~_- -
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5 "
en -
,-sl-5 , -
a y 2.. _
I I f
I y E J / f I / f I / f
/
/
1 . .
./
/ .
1 10 5 ' I / /
9 e-- / =/A _ __--z_ : _::- - :_-_X - --
g.5&el===l==:X- _T_T~ W ' - :- __
- - ^ - -
=
- 7. !T!=- /
- 6. l ,.
s.EIEl
,,. ___ ,s'._- . .- -_ _ _ _ - - - _ _ - . _ _ - _ _ . _ . -
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3.
--i
-I ^i'.-
-u
/ FIGURE 3 FAST NEUTRON FLUENCE (E>l MeV) AT 45' AS A FUNCTION
/ OF FULL POWER SERVICE (EFPY)
- 2. l J
/ !
! I
, i i
I i i #,
1 , , 22 ,
i , ,
i i I ! ! !
- i 17 I '
10 0 5 10 15 20 25 30 35 EFPY
CIRCUMFERENTIAL SEAMS VERTICAL SEAMS 19-8948 B6903-3 g e 10-894 8.4" 45'
' E CORE ,
CORE 1,44.0" .
B6903-2 19-894A l Y
~
C N -
L 20.1"11-894 B6919-2 20-894B 0 *45 CORE ,
3f ,
a j ,
4e 5 j
B6919-1 20-894A l
FIGURE 4 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE J.M. FARLEY UNIT 1 REACTOR VESSEL I
MANPTAL PROPERTY BASIS : , , ,
Controlling Matsrial : Low;r Shall (plate no. B6919-2)
Copper Content : 0.14 WT5 Nickel Content : 0.56 WIS c Initial RT NDT
- 3/47,121. 5'F Curve applicable for heatup rates up to 60 F/hr for the service period up to 16 EFPY and contains margins of 10 F and 60 psig for possible instrument errors 2500 , ,, ,
'. i l
l' LEAK TEST LIMIT- , ' , ' '
2250 . ,
j I I i ! !
I 2000 l
[ {
[
-- -e ! I 1750 'UNACCEPTABL E j j ACCEPTABLE OPERATION ,! ,i OPERATION _____
/
g 1500 /
- t. !
w 1250 HEATUPgATESUP / /
ly ,TO 60 F/HR (
i x /
y
[ 1000 \ / '
O
/ ' s y 750 ,
~ CRITICALITY LIMIT E
/ BASED ON INSERVICE
' / HYDROSTATIC TEST c
500 TEMPERAnlRE (287 7) lFOR THE SERVICE PERICO UP TO 16 lEFPY 250 0 300 350 400 450 500 0 50 100 150 200 250 IN01CATED TEhePERATURE (DEG.F)
Figure 5 Farley Unit 1 Reactor Coolant System Heatup Limitations Applicable for the First 16 EFPY
MATEDIAL PRDPEFTY BASIS :
Controlling Material : Lower Shell (plate no. B6919-2)
Copper Content : 0.14 h75 1 Nickel Content : 0.56 h75 l
Initial RT ET : 5F RT ET After 16 EFFY : 1/4T, 146.4*F
- 3/4T, 121.5'F
/hr for the servi:e Curvesapplicableforcooldownratesupto100{Fand60psigfor period up to 16 EFFY and contains margins of 10 possible. instrument errors 2500 l
I 2250 l
UNACCEPTABLE / ACCEPTABLE :::::
2000 OPERATION l OPERATION
/
1750
/
I g 1$00 /
G
- t. l y 1250 j E i 0 ,
l l 1000 f 8 _. i 3 750 :C00gDOWNRATES f F/HR 7 500_]ph ggs,fh
- 40'-35V 250 }}{ gh I
O ' O 50 100 150 200 250 300 350 400 450 500 IMOICATED TEWPERATURE (DEC.F) Figure 6 Farley Unit 1 Reactor Coolant System Cooldown Limitations Applicable for the First 16 EFPY
MATERIAL PROPERH BASIS : j Controlling Material : Lower Shell (plate no. B6919-2) Copper Content : 0.14 WT5 Nickel Content : 0.56,Wr5 Initial RT NDT
- 5F RT After 32 EFPY : 1/4T,160g NDT
- 3/4T, 132 F Curve applicable for heatup rates up to 60 F/hr for the service period up to 32 EFPY and contains margins of 10 F and 60 psig for possible instrtanent errors
-2500 ggigignimizini j
, 7 i
LEAK TEST LIMITS ' 1 2250 Nw 1
}
1 l l I r f 1
/
) i 2000 / /
i i ~~ UNACCEPTABLE i I ACCEPTABLE 1750 OPERATION 1
/ I
/ OPERATION --
I I es00 l l G I I
- 1250 ) i r
g r f id 1000 HEATUPgATESUP / E TO 60 F/HR \ --.
/
, f ,
< 750 : ,
Y
/ TRITICALITY LIMIT ,
M BASED ON INSERVICE E
/ HYDROSTATIC TEST 300 , TEMPERAWRE (300 F)
FOR THE SERVICE PERIOD UP TO 32 EFPY 250 0 150 200 150 300 350 400 450 500 0 50 100 INDICATED TEMPERAWRE (DEG. F) Figure 7 Farley Unit 1 Reactor Coolant System Heatup Limitations Applicable for up to 32.EFPY
MATERTAL PROPERTY BASIS : Controlling Material : Lower Shell (plate no. B6919-2) Copper Content : 0.14 W 5 Nickel Content : 0.5g W5 Initial RT NDT
- 5F RT After 32 EFPY : 1/4T,160g NDT : 3/4T, 132 F 0
/hr for the service Curves period up applicable to 32 EFPY forand cooldown contains rates up toof100 margins 10 {F and 60 psig for possible instrtanent errors 2900 ,
I f 2250 l I f 2000 ! 1750 UNACCEPTABLE ACCEPTABLE ~~~ OPERATION j OPERATION 2 --- I
/
g 1500 H / 2 v } g 1250 / D / i 0 A E 1000 / l b
< >[
S 750 y00gDOWNRATES M S ' m. F/HR .,,u i .->>// 500 --- 04 --cm /
- 202 l;;M/
~ 40" - -
/
250 --
-10060 'f 0 500 0 50 100 150 200 250 300 350 400 450 INDICATED TEMPERARJRE (DEG. F)
We8 Farley Unit 1 Reactor Coolant System Cooldown Limitations Applicable for up to 32 EFPY
MAHRTM PROPEPTY RISTS : . . . Controlling Material : Lower ShcIl (plata no. B6919-2) Copper Content : . 0.14 HTS Nickel Content : 0.56 h7% Initial RT NDT
- 5F
~ -
RT g After 16 EFPY : 1/4T,146.4* F
- 3/4T, 121.5*F~
Curve applicable for heatup rates up to 60 F/hr for the service period up to 16 EFPY 2500 ,,,,,,,,,,,,, , , , i I I iI I I i l i I iil I I I I l I i
- LEAK TEST LIMITS i 2250 h/ ; j l 1 I l , ,
2000 / / , UNACCErrABLE / / ! ACCEPTABLE 1 OPERATION j j OPERATION .-P-i
} }
/ /
G 1500 /
/
G HEATUPgATESUP S 70 60 F/HR \ > / w 1250 N/ ' i
; i 1000 f 8 - /
5 750 iRITICALITY LI ET E --- BASED ON INSERVICE
' HYDROSTATIC TEST#
500 TEMPERATURE Q.74 F) FOR THE SERVICE PERIOD UP TO 16 EFPY 250 0 150 200 250 300 350 400 450 500 0 50 100 IN01CATC0 TEWPERATURE (DEG.F) Figure 9 Farley Unit 1 Reactor Coolant System Heatup Limitations Applicable for the First 16 EFPY
i MATER ML PROPEFTY BASIS : Controlling Material : Lower Shell (plate no. B6919-2) Copper Content : 0.14 h75 Nickel Content : 0.56,h75 Initial RTg : 5F RTg After 16 EFPY : 1/4T,146.4*F
- 3/4T, 121.5'F Curves applicable for cooldown rates up to 100 F/hr for the service period up to 16 EFPY 2500 .
1 l r 2250 l 2000 , I i > ACCEPTABLE ! i-
" UNACCEPTABLE. i 1750 ' ' OPERATION ,
OPERATION i i i t l
' /
)
/
/
i
.. i
!1000 l
/
g --C00gDOWN RATES F/HR
~
U 750 - -
!i
OK ,
-w
- , 20e '
500 ---- 40 DM '
- ' 60&
--~~~1007 250 0 250 300 350 400 450 S00 0 50 100 150 200 INDICATED TEMPERATURE (DEG.F)
Figure 10 Farley Unit 1 Reactor C0clant System C00ldown Limitations l Applicable for the First 16 EFPY 1 i
MATERIAL PROPERTY BASIS : Controlling Material : Lower Shell (plate no. B6919-2) Copper Content : 0.14 WT5 Nickel Content : 0.5gWr% Initial RTg : 5F RT NDT After 32 EFFY : 1/4T,160g
- 3/4T, 132 F ,
Curve applicable for heatup rates up to 60 F/hr for the service period up to 32 EFPY 2500 iiiiiii i,,,, , , , LEAK TEST LIMIT s
' m
/
i I f i 2250 1 I 1 J J V I I I 1 2000 I
/ I
/
I i
- UNACCEPTABLE j j ACCEPTABLE - ----
IM OPERATION r I i r OPERATION : - J J I I 1500 G
- 8 i
}
r
}
k J J - 1250 / / Y ' o r
---- i m 1000 ---2 j a- ::
- HEATUP GATES UP TO 60 F/HR N ,'
2 g / ~ 5 H N / -CRITICALITY LIMIT
~~
BASED ON INSERVICE k ---- ' f HYDROSTATIC TEST 500 TEMPERA 1URE (288 F) FDR THE SERVICE PERIOD UP To 32 EFPY 250 0 0 50 100 150 200 250 300 350 400 450 Soo INDICATED TEMPERATURE (DEG. F) Figure 11 Farley Unit 1 Reactor Coolant System Heatup Limitations Applicable for up to 32 EFPY
4 MATERIAL PROPERTY BASIS : Controlling Materi'al : Lower Shell (plate no. B6919-2) Copper Content : 0.14 WT5 Nickel Content : 0.56 Wr5 Initial RT NDT
- 5F U
RTg After 32 EFPY : 1/4T,160g
- 3/47, 132 F Curves applicable for cooldown rates up to 100 F/hr for the service period up to 32 EFPY 2500 ,
j me l I 2000 UNACCEPTABLE r' ACCEPTABLE OPERATION . OPERATION i i 1750 I I _ 1500 / R i r 12
- a:1250 _
i l l g 1000 s
/
h {[? C00gDOWNRATES 5 750 F/HR y -- 0% m
,;;/,
soe
"~
20
- 40 3 gyj
-~
~ 60' e
- 100' 250 0 150 200 250 300 350 400 450 500 0 50 900 INDICATED TEMPERATURE (DEG. F)
Figure 12 Farley Unit 1 Reactor. Coolant System Cooldown Limitations Applicable for up to 32 EFPY
- - - . - -. . _._---- - ._.___. _ - - - ________. _- _. ___. .__ -.-_-}}