ML20005G421
| ML20005G421 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 02/28/1989 |
| From: | Ray N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20005G418 | List: |
| References | |
| MT-SMART-078(89, MT-SMART-078(89), MT-SMART-78(89, MT-SMART-78(89), NUDOCS 9001190134 | |
| Download: ML20005G421 (37) | |
Text
{{#Wiki_filter:,c l l Letter Report MT/ SMART-078(89) BYRON UNIT 2 AND BRAIDWOOD UNIT 2 REACTOR VESSEL HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION February 1989 4 Prepared by: .( N. K. Ray N '
) - (
Verified by: ,[[[1aw2/ko .
- 5. E. Jtnichko Approved by: 2 h I% K T. A. Meyert Manager Structural Materials Engineering Prepared for Commonwealth Edision Company 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 and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 900119o134 891117 - - _ ,, f.DR ADOCK 0500 3
1
- j. ,
l TABLE OF CONTENTS \ l ! Section Title Page ; 1.0. INTRODUCTION 1 q 2.0 FRACTURE TOUGHNESS PROPERTIES 1 3.0 ADJUSTED REFERENCE TEMPERATURE 2 4;0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE-RELATIONSHIPS3 ) f 5.0 HEATUP AND COOLDOWN LIMIT CURVES 6
6.0 REFERENCES
7 l
- APPENDIX A- HEATUP AND C00LDOWN DATA A.1 '
l 1 J
. s M .10 jj
i LIST OF TABLES i l Table Title Page 1 Byron Unit 2 Reactor Vessel Fracture Toughness Properties 9 ! 2 Braidwoon Unit 2 Reactor Vessel Fracture Toughness Properties 10 3 Calculation of Adjusted Reference Temperatures for Byron 11 Unit 2 Reactor Vessel Material - Circumferential Wald 4 Calculation of Adjusted Reference Temperatures for 12 Braidwood Unit 2 Reactor Vessel Material - Circumferential Weld LIST OF FIGURES Figure ) Title Page
;
1 Fluence Factor for Use in the Expression for ART 13 NDT 2 Byron Unit 2/Braidwood Unit 2 Reactor Coolant System Heatup 14 Limitations Applicable for the first 10 EFPY 3 Byron Unit 2/Braidwood Unit 2 Reactor Coolant System Cooldown 15 Limitations Applicable for the First 10 EFPY 4 Byron Unit 2/Braidwood Unit 2 Reactor Coolant System Heatup 16 I Limitations Applicable for the First 16 EFPY 5 Byron Unit 2/Braidwood Unit 2 Reactor Coolant System Cooldown 17 Limitations Applicable for the first 16 EFPY
HCATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION-
1.0 INTRODUCTION
Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperatureTTer the reactor vessel. The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture tough-ness properties and estimating the radiation-induced ART RT is NDT. NOT designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F. RT NDT increases 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, ART due to the radiation exposure associated with that NDT time period must be added to the originF unirradiated RT The extent of NDT. the shift in RT NDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulator Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)Il . Two sets of heatup and cooldown curves were generated based on the limiting material (2,3) in the Byron Unit 2 and Braidwood Unit 2 reactor vessel. The first set is for 10 EFPY and the second set is for 16 EFPY. 2.0 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plani 43. The pre-irradiation fracture-toughness properties for the materials in the Byron Unit 2 and Braidwood Unit 2 reactor vessel are presented in tables 1 and 2. an.. . 3
l-3.0 ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2 [1] the adjusted reference temperature (ART)
~
for each material in the beltline is given by the following expression: ART = Initial RTNDT + ARTNDT *i N""9 " (1) Initial RT NDT is the reference temperature for the unirradiated material as defined in paragraph NS-2331 of Section !!! 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 NDT by irradiation and should be calculated as follows: ART NDT = (CF)f(0.28-0.10 log f) = (CF) (ff) (2) The value, "f", used in equation (2) is the calculated value of the neutron fluence at the location in the vessel at the location of the postulated defect, n/cm2 (E > 1 MeV) divided by 10 10 . The fluence factor, "ff" is shown in figure 1. 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. I(depth X)
- Isurface (3) where x (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface. The attenuated fluence is then used in equation (2) to calculate ARTNDT at the specific depth.
CF (*F) is the chemistry factor, obtained from reference 1 for the beltline region materials of the Byron Unit 2 and Braidwood Unit 2 reactor pressure asavosoons to 2
E [A 'e 3 Q,
;
. vessel. The limiting material was found to be the circumferential weld for both Byron Unit 2 and Braidwood Unit 2. The calculation of ART for this limiting material is shown in tables 3 and 4. ART values at 1/4T and at 3/4T L locations will be enveloped between Bryon Unit 2 and Braidwood Unit 2. The enveloped ART _. values at 1/4T and 3/4T locations will be used to develop the reactor pressure vessel heatup and cooldown ' curves as described in the
- following sections. -
u 4.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various
.heatup.and cooldown rates specif.ies 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, l KIR, for the metal temperature at that time. K is obtained from the IR reference fracture toughness curve, defined in Appendix G to the ASME Code (5) , The KIR curve is given by the following equation: v KIR = 26.78 + 1.223 exp (0.0145 (T-RT NDT * + 160)] (4) where l' K!R = reference stress intensity factor as a function of the metal j temperature T and the metal reference nil-ductility temperature RT
- NDT I Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code (5) ,, f,)jo,,;
CKig + KIT
- KIR (5)
- NOTE: RTNDT as used in the ASME Code (5) is in fact the adjusted reference
- temperature (ART) as defined in NRC Regulatory Guide 1.99, Rev. 2 (1) and calculated in section 3.0.
3
4, ,
' ;
where. K!M = stress intensity factor caused by membrane (pressure) stress
' KIT = stress intensity factor caused by the thermal gradients [
u _KIR = function of temperature relative to ttis RT f the material NDT f 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 <
f At any time during the heatup or cooldown transient, K g is determined by tho' metal temperature at the tip of the postulated flaw, the appropriats value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calcu- -
~latedandthenthecorresponding(thermal) stress.intensityfactors,K IT' for the reference flaw are computed. From equation (5), the pressure stress
- intensity factors are obtained and, from these, the allowable pressures are y calculated.
For the calculation of the allowable pressure versus coolant temperature
- during cooldown, the reference flaw of Appendix G to the ASME Code (5) 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 ir.terest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. a n. m eeseae 4
I h - During cooldown, the 1/4 T . vessel location is at a higher temperature than the fluid adjacent to the vessel inside surface._ 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 KIR at the 1/4 T location for finite cooldown rates than for steady state operation. Furthermore, if conditions exist so that the increase
- in KIR exceeds KIT, 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 1/4 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 ensures conservative operation of the system for the entire cooldown period.
. Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature. relationships are developed for steady-state conditions as well as '
finite heatup rate conditions assuming the presence of a 1/4 T defect 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 f r the 1/4 T crack during heatup is lower IR than the K for the 1/4 T crack during steady-state conditions at the same IR coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIR 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 1/4 T flaw is considered. 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 the pressure-temperature limitations for the case in which a 1/4 T deep outside Nascososas to 5 '
surface flaw is assumed. Unlike the situation at the vessel inside surface,
=the thermal gradients established at the outside surface during heatep produce- ~
stresset,'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 musW 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 point-by point comparison of the p steady-state and finite heatup rate data.. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from [ the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling
+-
condition switches from the inside to the outside, and the pressure limit must ' at all times be based on analysis of the most critical criterien. - Finally, the 1983 Amendment to 10CFR50(6) has a rule which addresses 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 RT NDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure. 5.0 HEATUP AND C00LDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods described in section 4.0, and L the Westinghouse procedure of reference 7. 1 {. L Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in figures 2 G 6
n i f) and-3 for 10'EFPY and in figures 4 and 5 for 16 EFPY. This is in addition to
. other criteria which must be met before the reactor is made critical.
The leak limit curve shown in figures 2 and 4 represents the minimum temperature requirements at the leak test pressure specified by applicable codes [4,5) . The leak test limit curves were determined by the methods of references 4 and.6. Finally, tables 1 and 2. indicate that the limiting flange RT f 30'F NDT occurs in the vessel flange so the minimum allowable temperature of this
. region is 150*F.per reference 6.
These limits are less restrictive than the curves shown on-figurer,2 to-5. L Figures 2 to 5 define the limits for ensuring prevention of nonductile failure for the Byron Unit 2 and Braidwood. Unit 2 Primary Reactor Coolant System.
6.0 REFERENCES
h
- 1. Regulatory Guide l.99, Revision 2, " Radiation Embrittlement of Reactor ~
Vessel Materials," U.S. Nuclear Regulatory Commission, May, 1988. o
- 2. Response to V. S.-Nuclear Regulatory Commission Generic Letter 88-11 for the Byron Unit 2 Reactor Vessel, MT-SMART-217(88), November 1988. -
- 3. Response to U.S. Nuclear Regulatory Commission Generic Letter 88-11 for the Braidwood Unit 2 Reactor Vessel, MT-SMART-215(88).
- 4. " Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981.
- 5. ASME Boiler and Pressure Vessel Code, Section III, Division 1 -
Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure," pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York,1986. 7
, m- )
[ , H
;7 .
'6. Code of Federal Regulations, 10CFR50,-Appendix G, " Fracture Toughness.
i 4i t.. Requirements," U.S. Nuclear Regulatory Comission, Washington,: 0.C.,- I l Federal Register, Vol. 48 No.104, May 27,1983. u 7.. PProcedure.for' Developing Heatup and Cooldown Curves,"' Westinghouse Electric Corporation, Generation Technology Systems Divisi_on Procedure-GTSD-A-1.12 (Rev. 0), July' 13,1988. - i 4
?
. .;
[,
;
1 4 6 9 5 % wa w ,a m io 8
,e-TABLE 1 8YRON UNIT 2 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES Upper Shelf Energy Mat'1. Cu Ni- T RT NMWD(a) MWD (b)'
NOT NOT , Component Heat No. Spec. No. 1%J (%)- (*F) .(*F) (ft-lb). (ft-1b) i
~
, Closure head done C4375-2 A533 B, C1.'1 .12- .65 -40 -40 114 :--- Closure head ring 48C1300-1-1 A508 C1. 3 .05 . 69 -30 -30 -108 --- Closure head flange 2029-V-1 A508 C1. 2 ---
.71 0 0 157 ---
Vessel flange 124L556VA1 A508 C1. 2- ---
.70 30 30 129 ---
l Inlet nozzle 51-2979 A508 C1. 2 .07 .86 -10 -10 130. ---
Inlet nozzle 51-2979 A508 C1. 2 .07 .86 -20 -20 121 --- Inlet nozzle 42-5105 A508 C1. 2 .07 .84 .0 0 122 --- , Inlet nozzle 42-5105 A508 C1. 2 .07 .84 0 0 '121 --- Outlet nozzle 11-5052 A508 C1. 2 209 .85 -10 -10 108 --- e putletnozzle 11-5052 A508 C1. 2 .08 .81 -10 -10 121 --- Uutlet nozzle 4-2953 A508 C1. 2 .09 .78 -20 -20 133 --- * , Outlet nozzle 4-2956 .A508 C1. 2 .09 .81 -10 -10 121 ---- i Nozzle shell 4P-6107 A508 C1. 2 ---
.74 10 10 155 ---
Upper shell 490329/49C297'l-1 A508 C1. 3 .01 .70 -20 -20 149 149 Lower shell 490330/490298 1-1 A508 C1. 3 .05 .73 -20 -20 127 159 .; Bottom head ring 4801566 1-1 A508 C1. 3 .07- .67 -30 -30 126 ---
~ Bottom head dome C3053-1 A5338, C1. 1 .06 .64 -30 -20 121 -- . '
Upper shell to WF447 SAW (c) .059 .62 10 10 80 --- lower shell -i girth weld
' Weld HAZ --- --- --- ---
-60 -60 143 ---
(a) Normal to major working direction (b) Major working direction (c) Submerged arc weld 382Ss/040009 to
.~.--:.._- ....._k_..2a:__.:._-..-._ . - . - , - _ . - - . _ ,- _ _ _, _ _ _ _ _ _ . , _
m F. '!~ _ a TABLE 2' BRAIDWOOD UNIT 2' REACTOR VESSEL FRACTURE TOUGINIESS PROPERTIES Upper Shelf Energy mal'1. Cu Ni T WT . RT NT NWOI '} .NWD(b) 1 . Component Heat No. Spec. No.: '(%) -(%)- (*F) (*F)~ (ft-1b) jft-lb) Closure head done. B9754-1 A533 B, C1. 1 .16 .62 .-60 : -60 151 --- Closure head ring 50C478-1-1 A508 C1. 3 .05 .74 -30 -30 128 --- Closure head flange 2031-V-1 A508 C1. 2 ---
.75 20 20- 135- ---
Vessel flange 124P455 A508 C1. 2 .07 .70 20 20 128. --- Inlet nozzle 41-5414 A508 C1.'2 .07 .83 -10 -10 137 --- Inlet nozzle 41-5414 A508 C1. 2 .07 .84 -10 -10 140 Inlet nozzle 42-5417 A508 C1. 2 .09 .88 -10 -10 122 --- Inlet nozzle 42-5417 A508 C1. 2 .09 .88 -10 -10 116 E putlet nozzle 4-3502 A508 C1. 2 .09 .78 -10 -10 155 --- Dutlet nozzle 11-5226 A508 C1.'2 .09 .87 -10 -10 116 --- Outlet nozzle 4-3481 A508 C1. 2 . 07 .84 -10 163 --- Outlet nozzle 11-5266 A508 C1. 2 .09 .86 10 10 117 --- Nozzle shell SP-7056 A508 C1. 2 .04 .90 30 . 30 Upper shell 115 --- 49D963/490904 1-1 A508 C1. 3 .03 .71 -30 -30 119 147 Lower shell 500102/50C97 1-1 A508 C1. 3 .06 .75 -30 -30 -144
' Bottom head ring 4901066-1-1 168 A508 C1. 3 .07 .68 -30 -30 156 ---
Bottom head dome D1429-1 A5338, C1. 1 .11 .65 -20 -20 Upper shell to 120 --- WF562 SAW (c) .04 .67 40 40 80 --- lower shell girth weld Weld HAZ ---
-30 -30 '145 -- -
(a) Normal to major working direction (b) Major working direction l (c) Submerged arc weld ms.- i.
~m .am. . . . - . . _ -.._._m,~.m. _. -.1...._,__. .. _ . . . _ _ .
h
;-.
TABLE 3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING BYRON UNIT 2 REACTOR VESSEL MATERIAL - CIRCUMFERENTIAL WELD i, Regulatory Guide 1.99 - Revision 2 10 EFPY 16 EFPY 'i l= Parameter 1/4 T
~
3/4 T- 1/4 I 3/4 T
'ChemistryFactor,CF(*F) 80.6 80.6 80.6 80.6
- . Fluence,f--(10 19 n/cm)(a) 2 0.59 0.21 0.95 0.34 Fluence Factor, ff.. 0.85 0.58 0.98 0.70 L
ARTNDT = CF x ff (*F) 68.9 47.3 79.4 56.7 p ' Initial RTNDT, I ( F) 10 10 10 10 L Margin,M(*F) 56 47.3 56 56-
- - -
i Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 134.9 104.6 145.5 122.8
' ART = Initial RTNDT + ARTNDT + Margin (a) Fluence, f, is based upon fsurf(10 19 n/cm2 , E>l Mev) = 1.59 at 16 EFPY at inner surface. The Byron Unit 2 reactor vessel wall thickness is 8.50 ;
inches at the beltline region. 11
, ;-,,
f, TABLE 4'
- CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING BRAIDWOOD UNIT 2 REACTOR VESSEL MATERIAL -
;ja CIRCUMFERENTIAL WELD Regulatory Guide 1.99 - Revision 2 10 EFPY 16 EfPY
, Parameter 1/4 I 3/4 T 1/4 I 3/4 T ChemistryFactor,CF(*F) 54 54 54 54 Fluence,f(1018 n/cm2 )(a) -0.59 0.21 0.95 0.34 Fluence' factor, ff 0.85 0.58 0.98 0.70 i
********n****************************,...,****,n,....,**,,,***,,...,******,,,,, j ARTNDT = CF x ff ('F) 46.2 31.7 53.2 38.1
. Initial RTNDT, I ( F) 40 40 40 40 Margin,M('F) 46.2 31.7 53.2 38.1 i
******************,,,,,,,,,,,,u nen , ,, , ,, ,,,,,,,,,,,,,,,,,,,,,,,,,,,, ,
Revision 2 to Regulatory Guide 1.99 j Adjusted Reference-Temperature, 132.4 103.4 146.5 116.2 ART = Initial RTNDT + ARTNOT + iM*F9 "
.***,,,*****************,www*******************,****************,...........
19 1 (a) Fluence, f, is based upon fsurf (10 n/cm2 , E>l Mov) = 1.59 at 16 EFPY at inner surface. The Braidwood Unit 2 reactor vessel wall thickness is 8.50 inches at the beltline region. L I L o u
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l l MATERfAL PROPERTY BASIS i CONTROLLING MATERIAL - CIRCUMFERENTIAL WELD RTNDT_AFTER 10 EFPY: 1/4T, 134.9'F , 3/4T, 104.6'F p q, CURVES APPLICABLE FOR HEATUP. RATES UP TO 100*F/HR FOR THE SERVICE PERIOD UP L < ' 10 EFPY. CONTAINS MARGIN OF 10'F AND 60 PSIGJOR POSSIBLE INSTRUMENT ERRORS. 1 I 2500 u.m ,, , Leat lest l' , , , !- 2250 Limit i i
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for the Service -- 250 Period Up to 10 EFPY -: i ' o i i i i i iit , . 3 , . . . . ,' , 0 50 100 150 200 250 300 350 400 450 500 INolCATED TEMPERATURE (DCG.F) 1 Figure 2. Byron Unit 2/Braidwood Unit 2 Reactor Coolant System Heatup Limitations Applicable for the First 10 EFPY t 1 an..u io 14 1 _ . ._. _
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- CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD a RT !
NDT AFTER:10 EFPY: 1/4 T, 134. 9'F i L t- 3/4 T, 104.6'F l' 4 l' ! 1-b CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100'F/HR FOR THE S c TO 10 EFPY, ; P CONTAINS MARGIN OF 10'F AND 60 PSTrFOR POSSIBLE. INSTRUMENTi i
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O 50 100 '150 200 250 300 350 400 450 '200 INDICATED TEWRERATURC (DCCor) 1 ' Figure 3. Byron Unit 2/Braidwood Unit 2 Reactor Coolant System Cooldown Limitations Applicable for the First 10 EFPY an. one in 15
j' ;b
,3 MATERIAL PROPERTY BASIS.
;
~ CONTROLLING MATERIAli CIRCUMFERENTI AL- WELD -
RJNDT;AFTER16EFPY: 1/4T,146.5'F - 3/4T, 122.8'F a
~
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, 2,2 100 150 200 250 300 350 400 450 500
. INDICATED TEWPERATURC (DEC.F) t 1 p- Figure 4. Byron Unit 2/Braidwood Unit 2 Reactor Coolant System Heatup L Limitations Applicable for the First 16 EFFY ame. esome 16 l _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ = _ _ _ _ _ - - . _
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APPENDIX A-HEATUP AND COOLDOWN DATA 1t
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i .9 l CBE-CDEtOOF/HR HEATUP CURVE REG. GUIDE f . 99.RE V. 2 FOR WELD. TOE F PY - e02/24/89 { COtIPOSITE CURVE PLOTTED FOR HEATUP PROFILE 2 HEATUP RATE (S) (DEG.F/HR)'* 't00.0 IRRAOIATION PERIOD = 10.000 EFP YEARS FLAW OEPTH = (t-AOWIN)T - INDICATED 19mICATED INDICATED INDICATED IfelC4TED IDDICATED T ESIPERATURE ' PRESSURE (DEG.F) -(PSI)' TE9EPERATURE . PRESSURE '
-(DEG.F) (PSI) T E00PE RA TURE PRESSURE.
(DEG.F)- (PSI)- t 85.000 483.84 18 2 90.000 170.000 492.24 35 255.000 990.79 490.53- 13 175.000 505.48 3 95.000 ' 484.04 20 36 260.000 '1045.20 4 100.000 180.000 520.27 - 37 265.000 47f.22 21 185.000 536.56 ff03.52 5' 105.000 440.99 ' 22 190.000- 554.56 38' 270.000 =f166.35 6 910.000 452.70- 23 39 ' 275.000 1233.45 7 t15.000 195.000 574.50 '40 .280.000
.446.79 24 200.000 596.08 1305.74 8 120.000 442.76 41 285.000 .9382.43 9 25 205.000 619.71 42 125.000 440.64 26 310.000 S45.28
'290.000 1465.82 10 130.000 440,21 27 295.000 43 295.000 1554.44 91 135.000 672.97 : 44 300.000 1649.18 44i.47 28 220.000 703,15 12 140.000 444.25 29 225.000 45 305.000 1750,64 13 145.000 735.57 46 448.60 30 230.000 770.55 47 380.000 1858.87 14 150.000 454.40 31 235.000 305.000 1974.50 15 155.000 441.69
. 808.48 48 320.000 2097.99 32 240.000 949.19 49 16 160.000 470.42 33 245.000 325.000 2229.40' 17 165.000 480.55 893.00 50 330.000 2369.51
, 34 250.000 940.13 e
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CSE-CDE CDOLDOWN CURVES REG.IGUIDE 1.99,REV.2 FOR CIRC. WELD 10EFPY
' O2/2 /89 THE FOLLOWING DATA WERE PLOTTED FOR COOLDOWN PROFILEL1- ( STEADY-STATE COOLDOWN".)'
IRRADIATION PERIOD e 10.000 EFP ifEAR$ FLAW DEPTH'= 40 WIN T IteICATED . 18eICATED INDICATED - INDICATED TEMPERATURE ' PRES $URE.. TEMPERATURE PRESSURE IMICATED' 'IteICATED
- (DES.F) '<(PSI);' (DEG.F) (P5I)- TEMPERATURE : PRESSURE
'(DES.F)-
(PSI) i l 85.000 s 483.981 16 180.000. 6S7.78 l 2 90.000 490.53 17 31 235.000 1969.52
'3' 165.000 677.27' 32
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'S 905.000- ,K 993449 t 19 175.000 721.12' 34 20 100.000 745.86 250.000 1355.31 G t10.000 522.60 28 3S 255.000 1426.42~
185.000 771.80 36
'7 115.000 :532.05 _ . 22 190.000 000.13 . 37 260.000- 1502.76 '
8 120.000 S42.35 23 195.000 265.000 1984.53 9 125.000 830.37- 38 ( '583.42 - - 24 -200.000' 082.83 39 270.000 1672.16 l 10 130.000 565.32 25 205.000 275.000- +1748.06 it 135.000 '577,97 897.95 40 :290.000 '1866.65 12 26 290.000 635.54 di f40.000 591.72 27 215.000 975.89 295.000 1974.05 ! 13 145.000- SOS.St 28 42 290.000 2088.97 220.000 ' 14 f50.000 622.38 1099.22 43 295.000 22ft.88 29 225.000 1065.78 IS 955.000 639.34 44 ~ 300.000' 2343.06 30 230.000 tt15.88 45 303.000' '2482.95 l w 6 0 L I ..' 1 9 - Y _ --L w .. .N..,a.,.. .--,.u,,., N.---. -- s " ' '?" -- - - _ < u r,.-,m, _ , . , , ,,m.,,,., , y-,,, ,, ..m.
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~ IitRADI ATION PERIGO = ' 10.000 EFP YEARS-FLAW DEPTH = ADWIN T IpeICATED -INDICATED INDICATED INDICATED- INDICATED' IteICATED TEMPERATURE ,, PRESSURE -TEMPERATURE. PRESSURE'. . TEMPERATURE , PRE 55URE' -
(DES.F) '~(PSI)> ^(DEG.F) . 'tPST) (DES.F). 'tPSI). E f 85.000 443.76- ft 135.000 542.82- 21 :185.000 :749.55 2 90.000 450.70 12 140.000 557.40, 22 .190.000 779.64 2- 95.000 ~450.20 13 145.030 573.09 22 195.000 .8f2.20 .- 4 800.000 466.25 84 -150.000- 589.95 24 -200.000 347.08 5 105.000 474.95 IS 155.000 600.04 25 205.000 .884.62 . 6 110.000 484.20 16 . 160.000 627.43 26 210.000 924.90~ 7 815.000 '494.29 17- 165.000 648.49- 27- 2f5.000 968.23 . 8 120.000 505.13 18' 170.000 670.94 28 220.000 1014.79 - 9 f25.000 518.83 19 175.000 895.32 29 225.000 -1064.89. 10 130.000 ,
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CBE-CDE COOLDOWN CURVES REG. GUIDE f.99,REV.2 FOR CIRC.'WELO.f0EFPY 02/24/89 TifE FOLLOWING DATA STERE PLOTTED FOR COOLDOWN PROFILE.3.' ~ (40 DEG-F-/ HR COOLDOWN )- IRRADIATION PERIOD = 10.000 EFP. YEARS FLAW DEPTH = ADWIN T INDICATED INDICATED TEMPERATURE PitESSURE INDICATED . INDICATED INDICATED INDICATED' TEMPERATURE PRESSURE
-(DES.F) *'(PSI) (DEG.F)' (PSI) . TEMPERATURE PftESSURE (DEG.F)
(PSI) t 85.000 402.94- 135.000 2 90.000 490.19 if S07.56 20 f80.000 698.30 12 140.000 523.04 21 185.000
~3 95.000 'r 498.06 13 145.000 539.84 22 728.34 4 100.000 426.51 f4 f50.000 190.000 760.89 6 557.63 23- 195.000 105.000 435.58 f5 155.000 576.34 24 795.70 6 110.000 445.43 f6 200.000 833.f3 7 160.000 597.76 25 205.000 ff5.000 ass.00 17 f45.000 873.43 8 120.000 467.55 s20.24 26 210.000 Sis.77 9
18 170.000 644.31 27' 2f5.000 925.000 479.84 19 $75.000 G70.36 28 963.42 10 130.000 493.17 220.000 1013.58 e e W
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'CSE-CDE COOLDOWN CURVES REG. GUIDE't.99,REV.2 FOR CIRC. WELO.tOEFPY
'02/24/89 THE l'OLLOWING DATA WERE PLOT ~ED FOR COOLDOWN PROFILE 4 .(60 DEG-r / HR COOLDOWN - )
1RRAOIATION PERIOD
- 10.000 EFP YEARS FLAW DEPTH = AOWIN T --
IlmICATED- IDDICATED . INDICATED INDICATED . N: TEMPERATURE- C PitE55URE TEMPERATURE PRESSURE INDICATED- IteICATED (DES.F) ~M(PSI)' (DEG.F). ;(PSI)_ TEMPERATURE PitESSURE - (DEG.F) (PSI). 1- 85.000 361.36 2 11 135.000- 472.19 20 676.32 90.000 368.97 12 140.000 488.59- '21
'180.000-3 95:000 ~ 377.24 - 185.000 708.S t -
4 100.000 13 145.000 506.43 22. 190.000 743.60 386.08 14 150.000 525.64
'S 106.000 23 195.000- 789.07 6 110.000 395.75' 406.97 15 't55.000 546.27 24 200.000 821.39 '
16 160.'000 7 115.000 2417.45 568.59 - 25' 205.000 864.82 8 120.000 429.61 17 165.000' 592.57 26 210.000 911.56 = 9 125.000 18 170.000- 6f8.50 27 215.000 961.iss - 442.68 19 175.000 646.37 to 130.000 456.86 28 220.000 1016.03 0 0 e On l ._ y s 7 ..-A.g 9%.-s . . .---p,- . y m_. f., , y ..,p.% gy,,gg
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,. : 02/24/89 '+'
THE FOLLOWING - t- : DATA WERE CALCULATEDFOR THE IDtSERVICE HYOROSTATIC LEAK TEST.
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,n. REG. QUIDE, t.99 REV.2 FOR CIRC. WELD..tGEFPY. 02/24/09
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THE FOLLOtflNG DATA WERE PLOTTED FOR COOLDOWN PROFILE 2.. -(20 Ute-r / HR COOLootRt. ) IRRADIATION PERIOD &l'96.000 EFP YEARS FLAW DEPTH,* AOWI.N
+ , T,.
IPS)1CATED INDICATED INDICATED IISICATED IfeICATED ISSICATED T EGIPER ATURE ' PRES 5URE., T E8EPERA TURE PRESSURE TE8FE9ATURE PRESSURE (DES;F)' "(PSI)r'< (OES.F)" -(PSI) (DES.F) (PSI). t 95.000 < 429195? 12 -140.000- .824.82.. 22 -190.000 '7t2.50 2 90,000 434.99 13 145.000 537.95~ L23 195.000 739.90 3 05.000: R 44 9.19 ' '1 14 :-190.000 992.19 24 200.000, 789.301 4 100.000 .447.97 15 155.000 567.53 25 205.000 001.17 5 105.000'. C 489.29 ' te 180.000 -Se2.se 26 210.000 825.25 6 110.000 463.16 17 165.000 609.65 27 215.000 979.97 7 : 195.000
- f c479.65' te - '170.000~ '520.72' 20' .220.000 '99f.22' 8 120.000 480.68 19 175.000 649.17 29 ;225.000 953.75
't 125.000 490.53: '20 ' 480.000 883.20 30 230.000 999.26 10 130.000 501.13 21 185.000 606.93 31 235.000 1049.23 it t35.000 512.55 30 a O
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'THE.FOLLOtflNG DATA'tfERE PLOTTED FOR COOLDOWN PROFILE 3 (40 DEG-F / 95t.COOLDOWN. )'
IRRADIATION PERIOD C to.000 EFP YEARS-FLAW OEPTH = ADWIts T IMICATED ' ISICATED ISSICATED INDICATED IpeICATED INDICATED-iE00PERATURE. PRC$$URE. TEGEPERATURE PittSSURE , . T ESEPERA TURE" PRESSURE. (DEG.F)>- ' + '(PSI) -
~(DEG.F) .(PSI)- (des.F ) .(PSI) 1 ~88.000- 2397.20 . 12' 140.000 487.96 22 190.000 See.40 2 90.000 '383.26 f3 145.000 502.02 23 -195.000 . 717. 9 f .
3' 95.000 ' '399.84 14 .190.000- 517.15 24- 200.000 749.57 4 100.000 406.93 15 .155.000 533.36 25 205.000 793.59 8 906.000. '494.00: 98 '180.000 950.94 - 26. 280.000- ..820.18 6 f10.000 422.96 17 965.000 569.08 27- 295.000- 859.76 7 8t5.000' ~431.00 98 170.000 590.15 28 220.000 902.14 . 8 120.000 449.33 19 175.000 612.15 29 225.000 947.80
's $25.000 461.74 20 1e0.000 835.84' '30 230.000 996.81 90 $30.000 462.94 21 185.000 661.14 - 31 235.000 '9049.57 135.000
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