ML18012A453
| ML18012A453 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 12/30/1996 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18012A451 | List: |
| References | |
| NUDOCS 9701060061 | |
| Download: ML18012A453 (35) | |
Text
0 ENCLOSURE TO SERIAL: HNP-96-206 ENCLOSURE5 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RCS PRESSURE/TEMPERATURE LIMITS F I I P IFI TI PA E 970fObOObi 9bi230 PDR ADQCK 05000400 P PDR
2800 REACTOR COOt'ANT SYSTEM COOLDOV/N LIMITATIONS APPLICABLE UP TO 11 EFPY 2600 MATERIAL PROPERTY BASES:
Controlling Material = Plate A9153-1 2400 Copper Content 0.09 Nickel Content . o 6'.
Regulatory Guide Rev.
RTNpT Initial 60'F 2200 RTNBT at 1/4 T = 148'F RTNpT at 3/4 T = 133'F 2000 Pressure-Temperature Limits have NOT been adjusted'for instrument errors. These errors are contralled by the Technical Specification Equipment List Program, 1800 Plant Procedure, PLP-106.
1600 1400 P) 1200 C)
ABO 2O'F, SIN GLE CURVE FO R ALL RATES.
50'F /HR 1900 o C) 20'F/HR 800 IO'F/HR 5'F/HR 600 400 200 100 140 . 180 220 260 300 340 380 420 INDICATED TEMPERATURE DEGREES 'F FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPUCABLE UP TO 11 EFPY SHEARON HARRIS UNIT 1 3/4 4-35 Amendment No. 38
~ ~
, ~
2800 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 11 EFPY 2600 10'F/HR MATERIAL PROPERlY BASES:
Controlling Material = Plate A9153-1 00'F/H 2400 Copper Content 0.0 Nickel Content +4'/o Regulatory Guide Rev.
RTNpT Initial 60'F 2200 RTNpT at 1/4 T = 148'F RTNpT at 3/4 T = 133'F 2000 Pressure-Temperature Limits hove NOT been adjusted for instrument errors. These errors are controlled by the Technical Specification Equipment List Program, 1800 Plant Procedure, PLP-106.
1600 1400 P) 1200 C) 50'F/HR 1000 o+
O F/HR Z 800 20'F/ HR 10'F /HR 600 400 200 100 1 40 180 220 260 300 340 380 420 INDICATED TEMPERATURE DEGREES 'F FIGLIRE 3.4-3 REACTOR COOLANT SYSTEM HEATUP UMITATIONS APPUCABLE UP TO 11 EFPY SHEARON HARRIS UNIT 1 3/4 4-36 Amendment No. 38
REACTOR COOLANT SYST BASES SPECIFIC ACTIVITY Continued distinction between the radionuclides above and below a half-life of 15 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.
Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a-reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides.
The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about 1 day, about 1 week, and about 1 month.
Reducing T, to less than 500'F prevents the release of activity should a steam generator tube rupture occur, since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible the data obtained.
if justified by CE<>c 6 3 4.4.9 PRESSURE TEMPERATURE LIMITS 5eC'r.ion X>
The temperature an pressure changes during heatup and cooldown are limited to be consistent with' equirements given in the ASME Boiler and Pressure Vessel Code, Appendix G, and 10 CFR 50 Appendix G and H.
10 CFR 50, Appen ix G also addresses the metal temperature of the closure head flange and vessel flange regions. The minimum metal temperature of the closure flange region should be at least 120'F higher than the limiting RT NDT for these regions when the pressure exceeds 20K (621 psig for Westinghouse plants) of the preservice hydrostatic test pressure. For Shearon Harris Unit 1, the minimum temperature of the closure flange and vessel flange regions is 120'F because the limiting RT NDT is O'F (see Table 8 3/4 4-1).
The Shearon Harris Unit 1 cooldown and heatup limitations shown in Figures 3.4-2 and 3.4-3 and Table 4.4-6 are not impacted by the 120'F limit.
- 1. The reactor coolant temperature and pressure and system cooldown and heatup rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 and Table 4.4-6 for the service period specified thereon:
- a. -
Allowable combinations of pressure and temperature For specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and SHEARON HARRIS - UNIT 1 8 3/4 4-6 Amendment No. 38
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REACTOR VESSEL TOUGHNESS CHARPY INITIAL UPPER SHELF ENERGY HEAT Cu Ni Two< RT~~ TRANSVERSE COMPONENT GRADE NO ~jet.% ~jet.% ~oF ~F FT-LB Closure Hd. Dome A533,B,CLI A9213-1 -10 114 Head Flange A508, CL2 5302-V2 0 135 Vessel Flange 5302-V1 -10 -8 110 Inlet Nozzle 438B-4 . -20 -20 169 438B-5 0 0 128 I 438B-6 -20 -20 149 CO Outlet Nozzle 439B-4 -10 -10 151 4398-5 -10 -10 152 439B-6 -10 -10 150 Nozzle Shell A533B,CL1 C0224-1 .12 -20 -1 90 C0123-1 .12 0 42 84 Inter. Shell A9153-1 .09 -10 60 83 84197-2 PK.< . 0 -10 91 71 Lower Shell C9924-1 .08 -10 54 98 ff el C9924-2 .08 Pf.4q -20 57 88 Bottom Hd. Torus A9249-2 -40 14 94 Dome A9213-2 -40 -8 125 Weld (Inter &.Lower Shell O
Vertical Meld Seams) QP4184 P6 o5 . 91 -20 -20 >94 IA CO Meld (Inter. to Lower Shell 4-Girth Seam) 59&1 ) p4'oZ p5 q4 -20
-Bo Fnj Se5li <HnnhjjnlC'nIj~nnL njetjeL~ntn<c net" ~~-~jcA+e5". q,EWScF
1
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I s . I I
REACTOR COOLANT SYSTE BASES PRESSURE TEMPERATURE LIMITS Continued The cooldown and heatup limits of Figures 3.4-2 and 3.4-3 are based upon an adjusted RT>>y (initial RT>>, plus predicted adjustments for this shift in RT>>,
plus margin).
In accordance with Regulatory Guide 1.99, Revision 2, the results from the material surveillance program, evaluated according to ASTM E185, may be used to determine 5RT>>, when two or more sets of credible surveillance data are available. Capsules will be removed and evaluated .in accordance with the requirements of ASTH E185-82 and 10 CFR Part 50, Appendix H. The results obtained from the surveillan'ce specimens can be used to predict .future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The cooldown and heatup curves must be recalculated when the B,RT>>7 determined from the surveillance capsule exceeds the calculated hRT>>, for the equivalent capsule radiation exposure. C&hS,G 4ion Allowable pressure-temperature relationships for various cooldown an heatup rates are calculated using methods derived from Appendix G in of the ASHE Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM)
's technology. In the calculation procedures a semielliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The imensions of this po'stulated crack, referred to in ppen ix G of ASHE the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure.
To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RT>> is used and this includes the radiation-induced shift, QRT>> corresponding to the end of the period for which cooldown and heatup curves are generated.
The ASHE approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K,, for the Sec.bc~ Xi SHEARON HARRIS - UNIT I B 3/4 4-11 Amendment No. ~8
REACTOR COOLANT SYSTEM BASES PRESSURE TEMPERATURE LIMITS Continued metal temperature at that time. K, is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The K, curve is given by the equation:
K, = 26.78 + 1.223 exp [0.0145(T-RT~, + 160)] (1)
Where: K, is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RT~,. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C Ki+ Ka Kg (2)
Where: K, = the stress intensity factor caused by membrane (pressure) stress, K = the stress intensity factor caused by the thermal gradients, Kg = constant provided by the Code as a function of temperature relative to the RT~, of the material, Ab&
C = 2.0 for level A and 8 service limits, and C - I.5 for inservice leak and h drostatic (ISLH) test operation .
At any time during the heatup or cooldown transient, K, is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RT~ 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 factor, Q, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
COOLOOWN 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. Ouring 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.
SHEARON HARRIS - UNIT 1 8 3/4 4-12 Amendment No.38
REACTOR COOLANT SYS BASES PRESSURE TEMPERATURE LIMITS Continued heatup and the time (or coolant temperature) along the heatup ramp.
Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
Rather, each heatup rate of interest 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 cui ve 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 limita-.
tions because it is possible for conditions to exist such that 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.
The composite curves for the heatup rate data and the cooldown rate data in Figures 3.4-2 and 3.4-3 have not been adjusted for possible errors in the pressure and temperature sensing instruments. However, the heatup and cooldown curves in plant operating procedures have been adjusted for these instrument errors. The instrument errors are controlled by the Technical Specification Equipment List Program, Plant Procedure PLP-106.
"ISLH" pressure-tern erature (P-T cur es a be used o 'nservice leak and drostatic tests
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. ~hb gu~4;n 3h~~ucb ~ v~ss,
>QgQ r ice b &< 4WC Cek< &uS~M CanPL<4~ bed% 6+ ~'~ CcigiCal.
Althoug the pressurizer operates >n temperature ranges a ove those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.
LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or an RCS vent opening of at least 2.9 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than. or equal to 325'F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (I) the start of an idle RCP with the secondary water temperature of the steam generator less than 50'F above the RCS cold leg temperatures, or (2) the start of a charging/safety injection pump and its injection into a water-solid RCS.
The maximum allowed PORV setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the performance SHEARON HARRIS - UNIT I B 3/4 4-14 Amendment No. 38 (
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'I
2800 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 11 EFPY 2600 ISLH MATERIAL PROPERTY BASES:
Controlling Material = Plate A9153-1 2400 Copper Content 0.09K Nickel Content 0.46/o Regulatory Guide 1.99, Rev. 2 60'F 2200 RTNpT Initial RTNpT at 1/4 T = 148'F RTNpT at 3/4 T = 133'F 2000 Pressure-Temperature Limits have NOT been adjusted for instrument errors. These errors are controlled by the Technical Specification Equipment List Program, 1800 Plant Procedure, PLP-106.
1 600 LIJ 1400 P) 1200 Ci LIJ ABOVE 2 20'F , SING LE C URVE FO R ALL R ATES 1000 o+
50'F /HR O
20'F/HR 800 IOF/HR 5'F/HR 600 400 200 100 140 180 220 260 300 340 380 420 INDICATED TEMPERATURE DEGREES 'F FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLOOWN LIMITATIONS APPUCABLE UP TO 11 EFPY SHEARON HARRIS UNIT 1 3/4 4-35 Amendment No.
2800 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 11 EFPY 2600 10'F/HR MATERIAL PROPERlY BASES:
Controlling Material Plate A9153-1 00'F/HR 2400 Copper Content 0.09K Nickel 0,46K Guide 1.99, Rev. 2 Content'egulatory RTNpT Initial 60'F 2200 RTNpT at I/4 T = 146'F RTNpT at 3/4 T = 133'F 2000 Pressure-Temperoture Limits hove NOT been adjusted for instrument errors. These errors are contralled by the Technical Specification Equipment List Program, 1800 cn Plant Procedure, PLP-106.
1600 1400 P) 1200 C) 1000 o+
50F/ HR C)
F/HR z 800 20'F/HR IO'F/HR 600 400 200 100 140 180 220 260 300 340 380 420 INDICATED TEMPERATURE DEGREES 'F FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 11 EFPY SHEARON HARRIS UNIT 1 3/4 4-36 Amendment No.
REACTOR COOLANT SYSTE BASES SPECIFIC ACTIVITY Continued distinction between the radionuclides above and below a half-life of 15 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.
Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to erform the sampling. transpor t the sample, and perform the analysis of'bout
~
0 minutes. After 90 minutest the gross count should be made in a
~
reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides.
The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about 1 day, about 1 week, and about 1 month.
Reducing T, to less than 500'F prevents the release of activity should a steam generator tube rupture occur, since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will'e detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.9 PRESSURE/TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section XI, Appendix G. and 10 CFR 50 Appendix G and H. I 10 CFR 50, Appendix G also addresses the metal temperature of the closure head flange and vessel flange regions. The minimum metal temperature of the closure flange region should be at least 120'F higher than the limiting RT NDT for these regions when the pressure exceeds 20K (621 psig for Westinghouse lants) of the pr eservice hydrostatic test pressure. for Shearon Harris nit 1. the minimum temperature of the closure flange and vessel flange regions is -120'F because the limiting RT NDT is 0 F (see Table B 3/4 4-1).
The Shearon Harris Unit 1 cooldown and heatup limitations shown in Figures 3.4-2 and 3.4-3 and Table 4.4-6 are not impacted by the 120 F limit.
The reactor coolant temperature and pressure and system cooldown and heatup rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 and Table 4.4-6 for the service period specified thereon:
- a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and SHEARON HARRIS - UNIT 1 B 3/4 4-6 Amendment No.
TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS CHARPY INITIAL UPPER SHELF ENERGY HEAT Cu Ni RTm TRANSVERSE COMPONENT GRADE NO ~wt. I ~wt. R ~$ '( ~F FT-LB Closure Hd. Dome A533,B,CL1 A9213-1 -10 8 114 Head Flange A508, CL2 5302-V2 0 135 Vessel Flange 5302-V1 -10 -8 110 Inlet Nozzle 4388-4 -20 -20 169 438B-5 0 0 128 438B-6 -20 -20 149 Outlet Nozzle 439B-4 -10 -10 151 439B-5 -10 -10 152 4398-6 -10 -10 150 Nozzle Shell A533B. CL1 C0224-1 .12 -20 -1 90 C0123-1 .12 0 42 84 Inter. Shell* A9153-1 .09 .46 -10 60 83 B4197-2 .09 .50 -10 91 71 Lower Shell~ C9924-1 .08 .47 -10 54 98 C9924-2 .08 .47 -20 57 88 Bottom Hd. Torus A9249-2 -40 14 94 Dome A9213-2 -40 -8 125 Weld (Inter 5 Lower Shell Vertical Weld Seams)* 4P4784 .05 .91 -20 -20 )94 Weld (Inter. to Lower Shell Girth Seam)* 5P 6771 .03 .94 -80 -20 80
SHEARON HARRIS - UNIT 1 B 3/4 4-8 Amendment No.
REACTOR COOLANT SYST BASES PRESSURE/TEMPERATURE LIMITS Continued The cooldown and heatup limits of Figures 3.4-2 and 3.4-3 are based upon an adjusted RT >> (initial RT>> plus predicted adjustments for this shift in RT>>
plus margin .
In accordance with Regulatory Guide 1.99, Revision 2, the results from the material surveillance program, evaluated according to ASTN E185, may be used to determine hRT when two or more sets of credible surveillance data are available. Capsules will be removed and evaluated in accordance with the requirements of ASTM E185-82 and 10 CFR Part 50, Appendix H. The results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The cooldown and heatup curves must be recalculated when the hRT , determined from the surveillance capsule exceeds the calculated hRT>> for 5e equivalent capsule radiation exposure.
Allowable pressure-temperature relationships for various cooldown and heatup rates are calculated using methods derived from Appendix 6 in Section XI of the ASNE Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFN) technology. In the calculation procedures a semielliptical surface defect with a depth of one-quarter of the wall thickness, T. and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix 6 of ASNE Section XI as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and rovide sufficient safety margins for protection against nonductile failure.
o assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature. RT>>, is used and this includes the radiation-induced shift, ART>> corresponding to the end of the period for which cooldown and heatup curves are generated.
The ASNE approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heal;up or cooldown cannot be greater than the reference stress intensity factor, K<<, for the SHEARON HARRIS - UNIT 1 B 3/4 4-11 Amendment No.
REACTOR COOLANT SYSTE BASES PRESSURE/TEMPERATURE LIMITS Continued metal temperature at that time. K~ is obtained from the reference fracture toughness curve. defined in Appendix G to the ASME Code. The K curve is given by the equation:
K = 26.78 + 1.223 exp [0.0145(T-RT>> + 160)j (1)
Where: K is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temper ature RT.
Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C Krv + Kre s KiR (2)
Where: K, = the stress intensity factor caused by membrane (pressure) stress, K<< = the stress intensity factor caused by the thermal gradients.
K = constant provided by the Code as a function of temperature relative to the Ropy of the material, C = 2.0 for level A and B service limits, and C = 1.5 for inservice leak and hydrostatic (ISLH) test operations. (
At any time during the heatup or cooldown transient, K, is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for Ropy, 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 factor, K,, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
COOLOOWN 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. From these relations, composite limit curves are constructed for each cooldown rate of interest.
SHEARON HARRIS - UNIT 1 B 3/4 4-12 Amendment No.
REACTOR COOLANT SYST BASES PRESSURE/TEMPERATURE LIMITS Continued heatup and the time (or coolant temperature) along the heatup ramp.
Furthermore. since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
Rather, each heatup rate of-interest 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 con'sideration.
The use of the composite curve is necessary to set conservative heatup limita-tions because it is possible for conditions to exist such that 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.
The composite curves for the heatup rate data and the cooldown rate data in Figures 3.4-2 and 3.4-3 have not been adjusted for possible errors in the pressure and temperature sensing instruments. However. the heatup and cooldown curves in plant operating procedures have been adjusted for these instrument errors. The instrument errors are controlled by the Technical Specification Equipment List Program, Plant Procedure PLP-106.
"ISLH" pressure-temperature (P-T) curves may be used for inservice leak and hydrostatic tests with fuel in the reactor vessel. However, ISLH tests required by the ASME code must be completed before the core is critical.
Although the pressurizer'perates in temperature ranges above those for which there is reason for concern of nonducti le failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.
LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or an RCS vent opening of at least 2.9 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix 8 to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 325'F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than 50 F above the .
RCS cold leg temperatures, or (2) the start of a charging/safety injection pump and its injection into a water-solid RCS.
The maximum allowed PORV setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the performance SHEARON HARRIS - UNIT 1 B 3/4 4-14 Amendment No.
II 1
II
ENCLOSURE TO SERIAL: HNP-96-206 ENCLOSURE6 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REACTOR VESSEL SURVEILLANCE CAPSULE REPORT REVISION
~Bck~r~nl In compliance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program, Carolina Power & Light Company (CP&L) submitted a report to the NRC by letter dated April 2, 1992, documenting the results of testing surveillance capsule V. This report, BAW-2154, dated March 1992, (referred to herein as "the Appendix H" report or submittal) contained information related to the material properties of reactor vessel beltline materials. Also, in response to a May 13, 1994 NRC request to verify information contained in the NRC Reactor Vessel Integrity Database (RVID), CP&L submitted a letter dated June 10, 1994. Included in that response, CP&L provided a marked up corrected page 7-14 to the original Appendix H submittal of April 2, 1992, which revised some reactor vessel beltline materials End- Of- Life (EOL) Upper Shelf Energy (USE).
As a result of HNP 6-month response to Parts 2, 3 & 4 of GL 92-01, Revision 1, Supplement 1, submitted November 16, 1995, it is necessary to further amend some of the data contained in the Appendix H submittal referred to above. The November 16, 1995 letter indicated that as a result of the assessment, the best estimate chemistry has been revised for some of the beltline materials and that there were some minor material property changes for the dropweight temperature (T>>)
and the Upper Shelf Energy (USE), regarding weld heat 5P6771. The response also indicated that while these changes did not create an adverse impact to reactor vessel integrity, some documentation changes were necessary. Specifically, the response stated that a revision to CP&L's prior submittal of the 10 CFR 50, Appendix H results for the HNP would be necessary and that these changes would be submitted to the NRC during 1996.
Qhhng~
These changes are marked up on the attached pages (which include the prior June 10, 1994 corrections) and are to replace those of the original Appendix H submittal of April 2, 1992.
These changes are consistent with those changes proposed for the Technical Specification Bases 3/4.4.9 changes and affect the best estimate chemistry for the reactor vessel beltline materials, the unirradiated T>>r and USE for circumferential weld material 5P6771. As a result, the predicted EOL USE has also been affected.
These changes are acceptable because they do not create any adverse impact on reactor vessel integrity, for the reasons explained in CP&L's November 16, 1995 submittal in response to GL Page E6-1
C ENCLOSURE TO SERIAL: HNP-96-206 92-01, Revision 1, Supplement 1, and as explained in the Basis for the Technical Specification change with this submittal, and as re-enumerated below:
(1) Chemistry Data.
The detailed response contained in CF&L's November 16, 1995 letter indicated that as a result of the assessment, the best estimate chemistry has been revised for some of the beltline materials.
In some cases, the nickel content has increased, however, the associated chemistry factor remains the same. In other cases, either the nickel content or the copper content or both, have been reduced, and therefore the chemistry factor could also have been reduced or it remained the same. However, CP&L has elected at this time not to revise the docketed chemistry factors, and so there is no impact to the pressurized thermal shock reference temperature (RT~) or nil-ductility transition reference temperature (RT~>>) values.
(2) Test Data for Weld 5P6771.
The initial or unirradiated material properties of the reactor vessel beltline materials were originally based on test results presented in a report, WCAP-10502, "CP&L SHNPP Unit No.1, Reactor Vessel Radiation Surveillance Program", dated May, 1984. In particular, the tests results for the surveillance weldment material in the surveillance capsule program were performed by Westinghouse, except for the dropweight temperature test which was reported from weld procedure qualification tests conducted by Chicago Bridge & Iron (CB&I)Nuclear Company. The Westinghouse tests on the surveillance weld material were performed on a weldment that was fabricated by joining the two intermediate shell base metal plates A9153-1 and B4197-2. The weldment was fabricated using the same weld geometry, weld filler wire heat number 5P6771 and Linde 124 flux, lot ¹ 342 and is identical to that used in the actual vessel beltline circumferential (girth) seam. It also was subjected to a similar heat treatment (simulated stress relief and post-weld heat treatment) corresponding to that of the actual reactor vessel circumferential weld. Westinghouse also performed chemical analyses of the surveillance material. These test results were reported in CP&L's latest Appendix H report, previously submitted, April 2, 1992.
As indicated in CF&L's November 16, 1995 letter, newly acquired data has been obtained relating to the reactor vessel beltline circumferential weld. In particular, this data pertains to separate tests performed by CB&I (not included in the WCAP-10502) on another section of the same surveillance weldment that represented the circumferential weld as described above. These tests included dropweight testing for the weldment, but not a chemistry analysis.
(3) Dropweight Temperature for,Weld 5P6771.
Weld material 5P6771 was used in the reactor vessel circumferential weld joining the intermediate shell base metal plates (A9153-1 and B4197-2) to the lower shell base metal plates (C9924-1 and C9924-2) in the beltline region. The dropweight temperature, T>>, is currently Page E6-2
ENCLOSURE TO SERIA: HNP-96-206 stated in the Appendix H report to be -20'F. However, this is based on CB&I weld procedure qualification test data. The newly acquired additional test data that was obtained from the CB&I surveillance weldment tests provided information relative to the dropweight temperature, T~~,
which was determined to be -80'F. The heat treatment duration for the weld qualification test weldment was longer than that for the surveillance weldment. As such, the material properties between these weldments is expected to be different. Since the surveillance weldment heat treatment duration is more representative of the actual reactor vessel weld heat treatment, the T~~ test results from the surveillance weldment tests should be used in defining the dropweight temperature, T~r, for the reactor vessel circumferential weld. Therefore, the T~r value for the reactor vessel circumferential weld, SP6771, is being revised to -80'F.
(4) T>> z.,value for Weld 5P6771.
The unirradiated surveillance test data reported in the Appendix H submittal were obtained from WCAP-10502. The Westinghouse tests on a section of the surveillance weldment provided a series of Charpy V Notch (CVN) test results and related chemical analyses. The additional test data obtained by CB&I on the other section of the surveillance weldment also provided a series of CVN test results, but no chemistry analyses. The two sets of test results (Westinghouse and CB&I) differ slightly, but are not considered to be a significant difference. In general, the Westinghouse test results show a higher Charpy curve than from the CB&I test data. The CVN tests are used, in part, to determine the temperature at which the 50 ft-lb energy level is obtained (T>> .,), which is considered when determining the initial reference temperature for the nil-ductility transition (RT~r). The Westinghouse reported T,<<.,value, at O', for the surveillance weldment is at lower temperature than that for the CB&I test. The minimum T>> .value for the surveillance weldment, as determined by the CB&I tests, is now conservatively estimated at 40'F.
(5) RT~r value for Weld SP6771.
As described in the ASME Code,Section III, Article NB 2300, the initial reference temperature for the nil-ductility transition, RT>>>, is based on the higher of T~r or T>> .- 60'F, provided that 35 mils lateral expansion is also obtained at the T>> .,i, temperature. The RT~~ value for circumferential weld 5P6771, was previously reported in the Appendix H submittal to be -20'F based on a reported T>>r value of -20'F from CB&I weld qualification tests, which is higher than the T>> .,- 60'F value, at -60'F, based on the Westinghouse surveillance weldment tests.
Although the revised dropweight temperature, T~~ (-80'F), is lower (per CB&I surveillance weldment tests), this does not change the initial or unirradiated nil-ductility transition reference temperature, RT>>~, for weld 5P6771, as stated in the Appendix H report at -20'F. This is because the T>> z.value for the newly acquired CB&I surveillance weldment test data indicates a conservative T>> .value of 40'F, with at least 35 mils lateral expansion. Thus, the RT~~ is determined by the CB&I surveillance weldment Charpy 50 ft-lb (T>> .,) value, less 60'F, and remains at -20'F.
Page E6-3
ENCLOSURE TO SERIAL: HNP-96-206 (6) T3Q Q,,Q value for Weld 5P6771.
The CVN tests are also used, in part, to determine the temperature at which the 30 ft-ib energy level is obtained (T3Q Q g)), which is used to evaluate the change or shift in the reference temperature for the nil-ductility transition (RT~~) due to irradiation. Since the Westinghouse test results show a higher Charpy curve than from the CB&I test data for the unirradiated surveillance weldment material, the Westinghouse reported initial T3Q Q p) value is at lower temperature than that from the CB&I test. When irradiated surveillance material is removed from the reactor vessel and tested, the shift in the T,p a.,i, value will be larger using the Westinghouse initial T3Q Q,g) value versus the CB&I initial T3Q Q g) value. Therefore, the Appendix H test results for the evaluation of the surveillance capsule willprovide a conservative position, when the Westinghouse initial T3Q Q,}i, value is used. Ifthe weld chemistry factor is determined from the surveillance capsule results, based on RG 1.99, Revision 2, Position 2.1, a higher chemistry factor, which is conservative, will be obtained due to a larger shift in the T>> .value with the Westinghouse initial T,Q Q,JQ value than ifthe CB&I initial T~Q Q fQ value is used. (At this time, CP&L is not utilizing the surveillance results for the determination of the weld chemistry factor for weld SP6771, as previously discussed.) Thus, the Westinghouse test results for the surveillance weldment, as reported in CP&L's Appendix H submittal, remain valid for the purposes of evaluating the reactor vessel surveillance capsule program.
(7) Upper Shelf Energy for Weld 5P6771.
The Appendix H report includes the initial or unirradiated USE data for the surveillance weldment, for weld SP6771. For the surveillance weldment, the USE is reported as 92 ft-lbs, based on Westinghouse tests on a section of the surveillance weldment. The Appendix H report also includes the unirradiated USE data for the reactor vessel circumferential weld, 5P6771, and is currently reported as 88 ft-ibs (lowest of either the single or tandem weld wires). The 88 ft-lbs value was a typographical error and should have been stated as 85 ft-lbs, none-the-less, it was based on test results obtained by CB&I on a weldment for the weld procedure qualification tests.
However, as discussed above, the weldment provided for the weld procedure qualification tests received a longer heat treatment duration than both the surveillance weldment and the actual reactor vessel weld. The newly acquired CB&I test data for the other section of the surveillance weldment material, indicates that the initial or unirradiated Upper Shelf Energy (USE) for the reactor vessel circumferential weld, 5P6771, to be 80 ft-lb. Since the surveillance weldment test data is considered to be more representative of the actual reactor vessel weld than the CB&I weld procedure qualification test data, it is more appropriate to utilize the (lower) CB&I USE value of 80 ft-lbs for the reactor vessel weld material initial USE property, thus resulting in a more conservative determination of EOL USE value.
Although the newly acquired CB&I test data for the surveillance weldment material, indicates the initial or unirradiated Upper Shelf Energy (USE) for the reactor vessel circumferential weld, 5P6771, to be 80 ft-lb, this adversely affects the irradiated End-Of-Life (EOL) USE. However, the value will remain greater than the 50 ft-lb limit prescribed by 10 CFR 50, Appendix G, Page E6-4
0 ENCLOSURE TO SERIAL: HNP-96-206 paragraph IV.A.1. Specifically, the EOL USE is predicted to be 60 ft-ibs at the inside surface and 62 ft-lbs at the quarter thickness (T/4) location. This prediction is based on the Regulatory Guide 1.99, Revision 2, Position 1.2 method using a conservative percentage reduction in USE. Since this weld material is included in the surveillance program, it is expected that the percent reduction in irradiated USE would not be as great ifthe benefit of the surveillance capsule results were applied as allowed by RG 1.99, Revision 2, Position 2.2.
However, for the surveillance capsule program evaluation, the initial USE value for the surveillance weldment of 92 ft-lbs, based on the Westinghouse surveillance weldment tests, will continue to be used, as reported in the Appendix H submittal. This is because:
(i) The unirradiated USE for the surveillance weldment is compared to the irradiated USE of the surveillance weldment removed from the capsules and the shift or reduction in USE due to irradiation is determined. The Westinghouse surveillance weldment data on unirradiated material reported in CPkL's Appendix H submittal, provides for a higher initial USE of 92 ft-lbs versus that determined by the CB&I weldment test data of 80 ft-lbs. Thus, the shift or reduction in USE for the irradiated specimens is larger using the Westinghouse initial USE value stated in the Appendix H report, than with the CBEcl initial USE value. Therefore, the Appendix H test results for the evaluation of the reactor vessel surveillance capsule program willprovide a more conservative position, as the percent reduction in USE will be larger than ifthe CB8cl initial USE was utilized. Thus, iffuture predictions of the USE for the reactor vessel circumferential weld were to credit the results of the surveillance capsule program, a larger shift in USE will result in a lower EOL USE using RG 1.99, Revision 2, Position 2.2. Note that CPkL does not currently credit the results of the surveillance capsule program for the reactor vessel circumferential weld EOL USE determination since this material is non-limiting for the HNP reactor vessel. (For reactor vessel circumferential weld 5P6771, the EOL USE is determined on the RG 1.99, Revision 2, Position 1.2 method, using a conservative percentage reduction in USE, as described above.)
(ii) The Westinghouse surveillance weldment tests upon which the initial USE value of 92 ft-lbs is determined also included a chemical analyses. The copper content and the initial USE of this surveillance weldment are used in conjunction with the surveillance capsule fluence to determine the predicted reduction in USE of the surveillance weldment for comparison with the actual irradiated surveillance capsule USE reduction. (The CBAI weldment tests did not include a chemical analyses).
Therefore, for the Appendix H report, the initial or unirradiated USE of 92 ft-ibs for the surveillance weldment, willremain as stated for the evaluation of the surveillance capsule.
However, for the reactor vessel weld, an initial or unirradiated USE of 80 Mbs will be utilized.
Page E6-5
f e
ENCLOSURE TO SERIAL: HNP-96-206 (8) Upper Shelf Energy for Beltline Materials.
The reduction in USE, due to irradiation, for the HNP reactor vessel beltline materials is predicted using the RG 1.99, Revision 2, Position 1.2 method, which is based, in part, on the copper content. As stated above, the best estimate copper content for some reactor vessel beltline materials was reduced slightly. Namely, plate B4197-2, weld 4P4784 and weld 5P6771. A lower copper content results in a smaller percentage reduction in USE when using the Regulatory Guide 1.99, Revision 2, Position 1.2 method, which is the method applied to the HNP predicted EOL USE. However, the percentage reduction in USE due to the reduction in copper content for these beltline materials was conservatively maintained the same, (i.e. unchanged). Therefore the EOL USE values reported in the Appendix H submittal, as amended by the CP&L June 10, 1994 letter, remain the same, except for weld 5P6771, which was revised due to a change in unirradiated USE, as described above.
Qgnclu~i The changes to the Appendix H report do not affect any significant results relative to capsule testing nor neutron fluence, nor do they affect any conclusion of that report or change the withdrawal schedule for any surveillance capsules.
Page E6-6
b e -5. v t'o o e cto e se d-o - e act e o ess Cr'te o - S e o a 's U t Hater al1 Chemical Composition, End-of-Life RT~>, F~
w/o'" Ins)de T/4 Mall 3T/I Mall Reactor Vessel Heat Surface Location Location Initial Ins tde T/I Mall 3T/4 Mall Beltline Region Location Nuiber Type Copper Hlckel n/CII n/ca'" n/co'~ Rior ~ R Surface Location location Interaedlate Shell A9153-I SA533, Gr81 0.
a+4 9 3.42Et19 2.15Et19 8.47Etls 60 171 164 149 lnteriedlate Shell 84197-2 SA533, GrBI 0.50 3.42E+19 2.15Et19 8.47Et18 211 203 Lower Shell 0'+7 C9924-I SA533, Gr81 0.08 4448' 3.42Et19 2.15Et19 8.47Et18 54 155 150 Lower Shell 47 C9924-2 SA533, GrBI 0 3.42Et19 2.15Et19 8.17E+18 57 158 153 140 Intern. to Lower Shell (Meld) 0 O'3 Q~+
AB ASA/Linde IZI J40I 3.42Et19 Z. ISEtl9 8.47Etl8 . -20 107 101. 82 Oe b5 intern. Longit. Melds (Both) BC/BD ASA/Linde 124 4HN 0.91 1.33E+19 8.36Et18 3.29E+18 -20 125 114 93 lower Longlt. Melds (Both) o~S BA/BB ASA/Linda IZI 0.91 1.33Et19 8.36E+18 3.29E+18 -20 125 114 93 Nargln based on Regulatory Guide 1.99, Revision 2, and a standard deviation 0 for aeasured initial values of RT~,.
~Ail data per MCAP-10502, Hay 1984 CS oacACAdCL 'o ~lc <ch~~f Ce)L WC. ~S oaSSC.
Q~'lg4l'Calculated per Regulatory Guide 1.99, Revision 2, dated llay 1988
~section 6, Table 6-3 (this report)
Revised initial RT~~ value - see Paragraph 7.4.
ab e -6. val at'o o eacto esse d-o - ' ressu zed e ma S ock C te 'o - S ea o a 's U 't P
t Haterial Chemical Estimated PTS RG<6E Composition, Inside Inside Evaluation, F bl Surface Surface Reactor Vessel Heat EOL Fluence Initial End-of-Life Screening Beltline Region Location Number Type Copper Nickel n/cm'~ RTNore F'" 'RTog, F 32 EFPY Criteria Intermediate Shell A9153- I SA533, GrBI 0 '9 3.12E419 60 171 171 270 Intermediate Shell 81197-2 SA533, GrBI 0 '0 3.12E+19 91ldl 211 211 270 o.Q I Lower Shell C9921- I SA533 ~ GrBI . 0 ~ 08 3.12Et19 51 155 155 270 o+1 Lower Shell C9921-2 SA533, GrBI 0 F 08 J44I 3. I2E419 57 158 158 270 0'o& 0@~ 300 Interm. to Lower Shell (Meld) AB ASA/Linde 121 Ak9S 3.12EOI9 -20 107 107 o.o5 Interm. Longit. Melds (Both) BC/BD ASA/Linde 121 3496- 0.91 1.33E+19 -20 125 125 270 p o5 Lower Longit. Melds (Both) BA/BB ASA/Linde 12I JIA6 0.91 1.33Et19 -20 125 125 270
'll data per MCAP-10502, Hay 1981 ~
AS 0~~04<~ 4 n
C~isbg kghefcn>'neL pu'PfL MC ~G-Dc'/4
~Calculated per 10CFR50, Section 50.61'Section 6, Table 6-3 (this report)
Revised initial Roof value - see Paragraph 7.1
/
I ab e 7- va at 0 o eacto esse d-o - . e U e S e e - S ea o a s U t Material Estimated Chemical EOL Fluence Estimated EQL-U$ $
gE~c F compos/)on, Per R.G. 1.99/2 ) Estimated w/o Inside T/4 Mall Ini))~l EFPf to Reactor Vessel Inside SLfL1hs Beltline Region Location Heat Number Type Copper Nickel Surface n/cm".."
Locat)on n/cm " USE ft-lbs Surface T/4 Mall Location I.S. 1/4 O~~
Intermediate Shell A9153-I SA533, GrBI 0.09 3,42Etlg 2.15E<19 83 Cg~ C9' >32 >32 Intermediate Shell 0 &
84197-2 SA533, Gr81 0.50 3.42Etlg 2.15E<19 71 53 55 >32 Lower Shell C9924-I SA533, Gr81 0.08 3.42E+19 2.15EI19 98 99 P) >32 >32 Lo~er Shell C9924-2 SA533, GrBI 0.0 3. 42& 19 2. 15E 119 88 CC pf >32 >32 Inter. to Lower Shell Xeid 0 OS os+ 9o S>
AB ASA/Linde 124 3.42EI19 2.15E>19 >32 >32
)
Interm. Long it. Xelds (Both) BC/BD ASA/Linde 124 0.91 1.33E419 8.36E018 94 74 76 >32 >32 Lower Longi t. fields (Both) BA/BB ASA/Linde IZ4 0.91 1.33019 8.36E+18 94 74 76 >32 >32
'"All data per MCAP-10502, May 1984 aS omerdR 4 "best ~~~4e"~'i%r 8atiurnAek ~ @PAL ESg, 55-OOmg
~Upper-shelf values calculated per Regulatory Guide 1.99, Revision 2, dated May, .1988 "Section 6, Table 6-3 (this report)
~Revised initial upper-shelf energy value - see Paragraph 7.4 Iol ADD
'Calculated per Regulatory Guide 1.99, Revision 2, dated May 1988.
(f) 4 ~gecvaYiq< o 6'(o c~r cm4'nk u4ah aSSu~eL Qs &L EOL US< >4>~>'k'.
g ~
Q ~
.Q.~Q gyp Qc T~( q~J ~~ SP41ll (6fcdAQWti+ &CHAS)> p.C'OL ESZ 95-0 5
~ $~
~ . Table C-4. Charpy Impact Data From Unirradiated Meld Metal'est Absorbed Lateral Shear Temp, Energy, Expansion, Fracture, F ft-lb 10 in. Yo 350 97* 81 '100 250 96* 75 100 250 92* 79 100 160 92* 84 100 160 91* 78 100 90* 72 75 89* 73 97 78 64 90 40 75 60 85 40 65 58 80 52 45 65 0 46 41 55 0 42 36 50 J'
30 33 26 45
- 30 31 26 45
- 30 28 20 55
- 60 25 18 40
- 60 21 16 35
-110 15
-110 1 15
- Values used 'to determine upper-shelf ene value er ASTM E185-82 OC SUN<i. AGE Ch(Sv K 5IMCiM gaff(' ~gL Wp oL5-aa'Slk /le <(Q S'4)~
~4+ Qgwm'wa4'ion o(. grUL ~ckc <css<L we(d..egg C-5 8NsERvlcE cQMPANY
Ore C-4. Char Im act ata From Unirrad'ated Weld Meta 100 75
~ ~
50
~ 2 I ')5 0.10
~ 0.08 Ul g 0.06 0.04 Q 0.02 110 DATA SUIIVNY-100 T -20F HOT Trl (33 )a()
90 Trr (9) rr (a) a Trr (3() rr la)-
80
()32 4VG) 92 tt-)ba.
-20F NDT 2 70 60 50 ZJ 30 20 ttATtRtRt, l inde 124 weld 10 fLUENCK Hut No. 5P6771/0342
-200 -)00 100 200 300 400 500 oaks c3u(n(n((( t)e(taint'a 4() ~
Test Tettoerature, F
()n)r(ae4A g()(ue'h'Lan((( ca, 9()4 spec((u)er(. Refer b> c<)L 55(c l5 ()() 5 3 t (ueU (r)tt~ for ~
te~d. tr) khan Telactr)r welsa(L ctr(urn/(en>d ~d.
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