ML20155A323
| ML20155A323 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/28/1986 |
| From: | Congedo T, Mager T, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20155A277 | List: |
| References | |
| TAC-59902, WCAP-10910, NUDOCS 8604080364 | |
| Download: ML20155A323 (20) | |
Text
.,
WESTINGHOUSE CLASS 3
'ECAP-10910 l
Rev. 1 CUSTOMER DESIGNATED DISTRIBUTION
+
HEATUP AND C00LDOWN LIMIT CURVES FOR THE ALABAMA POWER COMPANY JOSEPH M. FARLEY UNIT 2 REACTOR VESSEL T. V. Congedo n
S. E. Yanichko T. R. Mager b
February 1986 M'
Approved:
5 J. N. Chirigos, Manager Structural 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 0604000364 060327 PDR ADOCK ODO 4
P 1449E:10/022886
1.
INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature). The most limiting RTNOT 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 RT is designated as the higher of either NOT.
NDT the drop weight nil-ductility transition temperature (TNOT) 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 increases as the material is exposed to fast-neutron radiation. Thus, NDT to find the most limiting RT at any time period in the reactor life, NOT ART due to the radiation exposure associated with that time period must NOT be added to the original unirradiated RT The extent of the shift in NOT.
RT is enhanced by certain chemical elements (such as copper, nickel and NOT phosphorus) present in reactor vessel steels. Westinghouse, other NSSS vendors, the U.S. Nuclear Regulatory Comission and others have developed trend curves for predicting adjustment of RT as a funcd on n uence and NOT 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)('}.
Regulatory Guide 1.99 was originally published in July 1975 with a Revision 1 being issued in April 1977.
Given the copper and phosphorus contents of the most limiting material, the radiation-induced ARI can be estimated from Figure 1.
Fast-neutron NOT fluence (E > 1 MeV) at the inner surface,1/4T (wall thickness) and 3/4T (wall thickness) vessel locations are given as a function of full-power service in Figure 2.
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 NOT*
l l
1449E:10/022786 -_
2.
FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture-toughness properties of the Farley Unit 2 reactor vessel materials are presented in Table I.
The fracture-toughness properties s
of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan $ )
The postirradiation f racture-toughness properties of the reactor vessel beltline material were obtained directly from the Farley Unit 2 Vessel Material Surveillance Program.
3.
FLUENCE CALCULATIONS For the purpose of revising heatup and cooldown curves for Farley Unit 2, which has limiting embrittlement characteristics in the intermediate shell course plate 87212-1, it is necessary to know vessel fast fluence (+ (E > 1 MeV)) at the azimuthal peak location. This peak location is at O' relative to the core cardinal axes, and at this angle, fast fluences are required at vessel inner radius, vessel 1/4T, and vessel 3/4T.
The calculations performed for this purpose consist of adjoint analyses, relating the f ast 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.
l The adjoint function was generated using the 00T discrete ordinates codeII and the SAILOR cross-section library I.
The SAILOR library is a 47 group, ENOF-8/IV based data set produced specifically for light water reactor applications.
In generating the adjoint function, anisotropic scattering was treated with a P expansi n f the cross-sections. The adjoint source 3
location was chosen along the inner diameter of the pressure vessel.
This calculation was run in R, e geometry to provide ~a power distribution importance function for the exposure parameter of interest (+ (E > 1 MeV)).
Having the adjoint importance function and appropriate core power distributions, the response of interest is calculated as R,0 " A I 1(R,e) F(R,e) R dR de R
R e 1
1449E:10/022886 l
whero:
R Resp nse f interest (+ (E > 1.0 MeV), dPa, etc.) at radius
=
R,e R and azimuthal angle e.
I(R,9)
Adjoint importance function at radius R and azimuthal angle
=
e.
F(R,e)
Full power fission density at radius R and azimuthal angle
=
l e.
1 It should be noted that as written in the above equation, the importance l
function 1(R,9) 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 2 were taken f rom the following Westinghouse fuel cycle design reports for each operating cycle to date:
Fuel Cycle Report i
WC AP-10187 l
3 WC AP-10410 l
4 WCAP-10674 Of these, Cycles 1 through 2 utilized out-in fuel loading patterns, and Cycles 3 an. 4 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 DOT transport analysis performed in support of WCAP-10425, " Analysis of Capsule U 1449E:10/022886,
from the Alabama Power Company, Joseph M. Farley Unit 2 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 0*
2 Lifetime (n/cm )
EFPY Vessel IR Vessel 1/4T Vessel 3/4T 18 18 II 1.09 2.027 x 10 1.200 x 10 2.789 x 10 18 I8 lI 1.86 3.625 x 10 2.147 x 10 4.989 x 10 18 18 II 2.95 5.527 x 10 3.273 x 10 7.606 x 10 18 18 lI 3.20 5.898 x 10 3.493 x 10 8.116 x 10 II I9 18 32.0 4.999 x 10 2.960 x 10 6.878 x 10 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 f actor, K, for the combined thermal and pressure stresses at any time during heatup g
and cooldown cannot be greater than the reference stress intensity f actor, Kgg, for the metal temperature at that time.
K, is obtained from the g
reference fracture toughness curve, defined in Appendiu G to the ASME Code (5)
The K curve is given by the equation:
gg l
Kgg = 26.78 + 1.223 exp (0.0145 (T-RTNDT + 160))
(1) where K;g 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 NOT.
defined in Appendix G to the ASME Code (5),3 g,jjogg CKgg + Kit I "IR (2) 1449E:10/022886 where:
K, is the stress intensity f actor caused by membrane (pressure) stress g
i K
is the stress intensity f actor caused by the thermal gradients it K
is a funtdon of Wwatum niadve to the RT f the material gg NOT 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.
At any time during the heatup or cooldown transient, K is determined by IR 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 f actors are obtained and, f ro:n 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.
l 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.
j 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 1
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 conditiori, of course, is not true for the steady-state 14490:10/022786 _
situation.
It follows that, at any giv:n reactor coolant temp rature, the AT developed during cooldown results in a higher value of K, at the 1/4T g
location for finite cooldown rates than for steady-state operation.
Furthermore, if conditions exist such that the increase in K exceeds IR Kit, the calculated allowable pressure during cooldown will be greater than the steady-stnte value.
The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the 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 for the 1/4T gg crack during heatup is lower than the K f r the 1/4T crack during IR 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 Kgg'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 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 pressure-temperature limitations for the case in which a 1/4T deep outside 1449E:10/022786._ _
surface flaw is assumed. Unliko the situatico at the vassal insida 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 depenoent 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 is possible for conditions to exist 6.herein, 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 3 and 4.
In addition, heatup and cooldown curves without instrument errors are presented in Figures 5 and 6.
Based on the Farley Unit 2 fracture analysis results from Reference 6, the heatup curves in Figures 3 and 5 and the cooldowa curves in figure 6 are impacted by the new 10CFR$0 rule.
This rule states that the minimum metal temperature of the closure flange regions should be at least 120'F higher than the limiting RT for these regions when the pressure exceeds 20 percent of NDT the preservice hydrostatic test pressure (621 psig for Westinghouse plants).
However, the cooldown curve in Figure 4 is not impacted by the rule. Since there are many conservatisms (safety factor of 2 on pressure, K toughness g
and 1/4T flaw) built into the ASME Appendix G analysis method
, 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.
1449E:lD/022786 I
An evaluation has b:en performed to d2termine the acceptability of the Overpressure Mitigatio. System (OMS) presently in Farley Unit 2 (Technical Specification 3/4.4. Q.3) with respect to the 8 EFP't Heatup and Cooldown curves shown in Figures 5 and b respectively. For the purpose of the evaluation it was assumed that the residual heat removal,(RHR) relief valve lifts at 495 psig which includes 10% accumulation. The Heatup curve in Figure 5 does not fall below 495 psig at any temperature. A comparison to cooldown curves in Figure 6 shows that in the low temperature range (<130*F) cooldown j
rates of 20*F/Hr and lower fall above 495 psig. Although the cooldown curves for rates of 40*F/Hr and above do fall below 495 psig, it is not expected that the Appendix G curves will be violated during an actuation of the OMS since l
cooldown rates greater than of equal to 40'F/Hr are highly unlikely at low 7
l temperature conditions. Therefore, the Appendix G curves as illustrated in
[
Figures 5 and 6 will not be violated as the result of an actuation of the CMS.
i I
S.
HEATUP AND C00LDOWN LIMIT CURVES 1
l
\\
l Limit curves for normal heatup and cooldown of the prinury Reactor Coolant L
System have been calculated using the methods discussed previously, The derivation of the limit curves is presented in the NRC Regulatory Standard l
Review Plani 2I.
I Transition temperature shifts occurring in the pressure vessel materials due j
to radiation exposure have been obtained directly f rom the reactor pressure
)
vessel surveillance program.
l i
i l
Allowable corabinations 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 3 and 5.
This is in addition to other criteria which must be met before the reactor is made critical.
l The leak test limit curve shown in Figures 3 and 5 represent minimum temperature requirements at the leak test pressure specified by applicable i
codes (2,5) i i
I 1449E:10/022886 l
6.
AVAILABLE SURVEILLANCE CAPSULE DATA 18 2
Charpy test specimens f rom Capsule U irradiated to 5.61 x 10 n/cm indicate that the representative core region seld metal and limiting core region shell plate 87212-1 exhibited maximum shifts in RT 0*F and NOT 133*F, respectivelyI I.
The shell plate shift of 133*F is less than the 157*F shift which is predicted by Regulatory Guide 1.99 Revision 1.
The ART s used to compute the heatup and cooldown curves were obtained NOT from the radiation damage curve associated with the surveillance shell plate 18 2
shift of 133*F at 5.61 x 10 n/cm shown in Figure 1.
7.
SURVEILLANCE CAPSULE REMOVAL SCHEDULE The surveillance capsule withdrawal schedule for Unit 2 (Table !!) should l
remain the same as identified in the Technical Specifications and i
WCAP-10425( }. The dosimetry analysis of the second capsule to be removed af ter 4 EFPY should be used to re-evaluate the withdrawal schedule for the remaining capsules.
I i
i l
1449E:10/022786 - - - _.
REFERENCES (1) Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Raciation Damage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, April 1977.
(2) " Fracture Toughness Requirements," Branch Technical Position - MTEB No.
5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Saf ety Analysis Reports for Nuclear 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 OLC-76 " Coupled, Self-Shielded, 47 l
Neutron, 20 Ganna-Ray, P, Cross Section Library for Light Water 3
Reactors, Radiation Shield Information Center, Oak Ridge National Laborato ry.
l (5) ASME Boiler and Pressure Vessel Code, Section 111, Division 1 -
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.
(6) Miller, J. C., " Response to NRC Comments on Farley Unit 2,* ALA-85-706, July 31, 1985.
(7) Kurika, M.
K., Yanichko, S. E., Cheney, C. A. and Kaiser, W. T. " Analysis of Capsule U f rom the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program " WCAP-10425, October 1983, 1
\\
1449E:10/022786 -
TABLE I FARLEY UNIT 2 REACTOR VESSEL TOUGHNESS DATA Average Upper Shelf Enerew Normal to Principal Principal Working Working Cu P
N1 T
RT Direction Direction NOT NOT Component Code No.
Grade m
M m
{*F1
{ *g, (ft-lb)
(ft-lb)
CL. HO. Dome 87215-1 A533,8,CL.1 0.17 0.010 0.49
-30 16(a) 83(a) 128 CL. HD. Flange 87207-1 A508,CL.2 0.14 0.011 0.65 60(a) 60(a)
>56(a)
>86(C)
)
VES. Flange 87206-1 A508,CL.2 0.10 0.012 0.67 60(a) 60(a)
>7)(a)
>i09 Inlet Noz.
87218-2 A508,CL.2 0.010 0.68 50(a) 50(3) 103(a) 158 0.010 0.71 32(a) 32(a) 112(a) 172 Inlet Noz.
87218-1 A508,CL.2 Inlet Noz.
87218-3 A508,CL.2 0.010 0.72 60(a) 60(3) 98(a) 150 0.010 0.73 60(a) 60(a) joo(a) 154 Outlet Noz.
87217-1 A508,CL.2 Outlet Noz.
87217-2 A508,CL.2 0.010 0.72 6(a) 6(a) jog (a) 167 Outlet Noz.
87217-3 A508,CL.2 0.010 0.72 48(a) 48(a) jo3(a) 158 0.010 0.73 30 30(a) 97(a) 14g Upper Shell 87216-1 A508 CL.2 Inter Shell 87203-1 A533,8,CL.1 0.14 0.010 0.60
-40 15 99 140 l
Inter Shell 87212-1 A533,8,CL.1 0.20 0.018 0.60
-30
-10 99 134 tower Shell 87210-1 A533.8,CL.1 0.13 0.010 0.56
-40 18 103 128 Lower Shell 87210-2 A533,8,CL.1 0.14 0.015 0.57
-30 0
99 145 Trans. Ring 87208-1 A508,CL.2 0.010 0.73 40 40(3) 89(a) 137 801. HD. Dome 87214-1 A533.8,CL.1 0.11 0.007 0.48
-30
-2(a) 87(a) j34 Inter. Shell A1.46 SMAW 0.02 0.009 0.96 0(3)
O(a)
>131 Long Seams A1.40 SMAW 0.02 0.010 0.93
-60
-60
>106 Inter Shell to Lower Shell G1.50 SAW 0.13 0.016
<.20(b)
-40
-40
>102 Lower Shell Long Seams G1.39 SAW 0.05 0.006
<.20(b)
-70
-70
>126 (a) Estimate per NUREG 0800 "USNRC Standard Review Plan
- Branch Technical Position MTE8 5-2.
1 (b) Estimated.
j l
(c) Upper shelf not available, value represents minimum energy at the highest test temperature.
1
TA8LE II SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following removal schedule is recommended for future capsules to be removed from the Farley Unit 2 reactor vessel:
Lead Estimated Fluence Capsule Factor Removal Time [a]
n/cm2 x 1019 U
3.12 Removed (1.1)
.56 (Actual)
W 2.70 4
2.18 X
3.12 6
3.78(b)
ICI Z
2.70 12 6.54 V
3.12 18 11.34 Y
2.70 Standby (a] Effective full power years f rom plant startop (b) Approximates vessel end of life 1/4 thickness wall location fluence (c] Approximates vessel end of life inner wall location fluence 1449E:10/022886,
A = [40 + 1000 (% Cu - 0 08) + 5000 (% P - 0 008)l [f/10) 1/2 OPPER u
b200 5
e o#
=
?g O
$104 y
g t:
a O
0 50 nu 0 35 0 30 0 25 0 20 % Cu O 15% Cu 0.10% Cu o
%P = 0 012 LOWER LIMIT 5
~
% Cu = 0 08 y
% P = 0 M8 I
O Plate 07212-1 1
t i i i I
e t i a e i
a e
i 3 a 3
e a a ie a
R l
3 3 3 3
2a10" 4
6 3
10
2 4
6 8
10
2 4
6 FLUENCE, n/cm'(E > 1 Mev)
Figure 1.
Predicted Adjustment of Reference Temperature,"A", as a Function of Fluence and Copper Content.
For Copper and Phosphorus Contents other than those Plolled, use the Expression for "A" given on the Figure.
__ \\
r 2
10.0 9..
' ^ ' ' '
f g,,,
- <t.
j
. l.
. _.i.
..=:. =._. M-1 :-l
-E-liu. } +_i_.:
.a+i
'_ J _; a c.: 2.;_-iEM :r-.
.iHl-y~ ~
t-6.
.*
- 1..._.. '......... _..
... g..
'- ~
..... 3..
5
- *~-
~ ~ - - - * - -
~~
-' ~..
Rft
-4:.'..":
6.
W4 2. +i----a 4 - +f ' ' RFACE i
4...
i
' I
~
I
- 4.
I!
". d -.-
l i
1 3
- .4.
L2
. J.-
...,_ -. -.: - --. :. g 1/4T
.._7.....
..t 4.
... l r
....s.
l 2...
.__.4_
a
_4.
1
.4..
s t
=
l f_.
..n..v
_ _ L __..
'p 1
/
1 10 9--.... ; s. s
.f s.
4 _
i i
/
i 9,,
- s.
y,/
~
- s. -
i i
i
-r.
s.
8..
-. M.- p8-i-f-i nt :fr :--s :-- t -i " - ' --- M :: 1 -- -
i it a -i-4-d.
. r __ -
.s -i
-~
7..
--.:.-fj - / : : l i f. i.if
-: i- :
2 N i -- - - i i --j :-i-- l 2.:r.: j i s-
}.:
--l
..__.4. _-.
6
....t
- :.4 :
i
._._.L_.:....2_.4_..
.. Jf s.. -.
4 N
=.. - - i r_i.-
2._._2_.
. :.u...
. t:
3,
- e....
r.. _. J u. c. t e _ e 8
.4 :
~
%.4 f:. e.T/n 51.m.44_-.-._ _ 2 ~. t. -
- - - I f. Le di+e r e m h-2 5 : t b'
_ri-s / 1 /" N M.[3 5--D:
4 i --'I ~ ' I r
/-
- h ISE-- I 5~i 0 1
Mil: - i i '~" - " ~~~ -h
. u.s
... -_. _ _.. e.; : - f.
_.a.
.n. =_;_t.::
y
.,._.........$........i.
m
_ ~._.
a
- s. _
.s.
--.e
. -..._.....- j m.
......_4,,.__.._4
...c..
{
2..
...j g
g 4-
._7...
. _..q 3
.~ _.......
f.4_...
.__.L.
-.__.4.-.
w 3
I.
..g.q
__ _ ~..
.g
~. _.. _..
18
_i I
i 10 I
t I
g W e.:. j. ;f -. f. + -j -. : -r -
..- r. 4 i-0 3 -.+.
-. rt --
r ---.
.m-
.1 g
4H-M4 / ).-~b M.+3 4-7+ 4 2-1 Oi.7t.4;-
" 3 M f %-- - :t=4 iTM c.2 :*w*
=2
- == ':.
$ 3 ::- Y 4 5b--'_~-! 5 I-M?:#fs ) *Ml-5. -. -N 9.? 'IlMiNCii55N ' " 5.25 7.
..........._..._..a_.._ _. _.... _ _ _
6.
y
. _. _...._=..:_._. _
- f t
i a
_ s r.d Hs
- 4
{.. :-
f; rmuRL 4 l PAsi RtumVN FLUtEL (Es1 Mel) A5 A FUNCTION OF FUi.l. POWER i f :i
.. j3LRVRt L i t,t. ( r.t r'T ) : :.. ;
t...j :.1-. -
..-.. L.;..i. J)...;
6 3
_2._._.
.__1 t......
_L....._.. _..._.
t.
._ _1 -
1..
2.
_ _. =..
1
=-
1 i
..- [_.._
y 10 '
o a
1O lb 10 25 30 35 EFPY
MATERIAL PROPE1<11 RAMTM CO EROLLING MATERIAL :
R. V. IhTERMEDIATE SHEI.L COPPER C0hTER
- 0.20 WT1 PHOSPHORUS CONTENT
- 0.018 WT%
INITIAL RT
- . jot gyg.
RT AFTER 8 EPY
- 1/47, 146 F NDT
- 3/4T, 83 F CURVES APPLICABLE FOR HEATUP RATES UP TO 60"F/MR FOR 1HE PERIOD UP TO 8 EFPY AND CONTAINS MARGINS OF 107 AND 60 INSTRUMEhT ERRORS n....
r ll J
J 1.EAx TEST uxIT x
/
/
N /
f f
(
)
I
. u....
t
)
(
(
UNACCF.nABLE
/
/
- OPERATION.
(
(
~
L
=
I
/
/
i E
HEARIPRATESUP
[
[
{
TO 607/HR N
[
[
1
.h
/'
/
1 j
ACCEPTABLE OPERATION r
r r
e._<
j o
/
f
,,,, CRITICALITY LIMIT
/
/
/
'~
BASED ON INSERVICE HYDROSTATIC TEST
('
TEMPERATURE (287"F) f FOR THE SERVICE PERIOD UP TO 8 ETPY 3,
i mi smia mmmM l 'I
]
='
- a...
- n...
IEDICATED TEMPERATURE (DES.F)
FIGURE 3 FARLEY UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 8 EFPY MATERIAL PROPERTY BASIS CORTROLLING MATERIAL :
R. V. INTERMEDIATE SHELL COPPER CONTENT
- 0.20 WT5 PFOSPHORUS CONTENT
- 0.018 UT%
INITIAL RTET
- -10%
RT AFTER 8 EFPY 1/4T, 146 F ET
- 3/4T, 83 F CURVES APPLICABLE FOR COOLDCMN RATES UP TO 100"F/HR FOR 1EE SERVIC PERIOD UP 10 8 EFPY AE C0hTAINS MAEINS T 10'F AND 60 PSIG FOR POSSI INSTRUMENT ERRORS sees.e f
l
/
i
- ,e...e
/
e
/
r E
UNACCEPTABLE g
OPERATION
/
r
(
/
l
/
ACCEPTAB*.E e
, sees.e f
,0PERATION
~~--
f M
Om
$Yl 20 5::W go #
s-60
- 100 ',,
e.e e.e see.e see.e see.e 4ee. e' see.e IEDICATED TEMPERATURE (CEG.F) i FIGURE 4 FARLEY UNIT 2 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE FOR THE FIRST 8 EFPY 16 -
L
MATERIAL PROrkim BASIS _
CONTROLLING MATERIAL :
R. V. INTERMEDIATE SHELL COPPER CONTENT
- 0.20 Yr5 PH0SPHORUS CONTENT
- 0.0},8WT%
INITIAL RT
- -10T g
- 1/4T, 146 F
- 3/4T, 83 F CURVES APPLICABLE FOR HEATUP RATES UP TO 60 FMIR FOR EE SERVICE PERIOD UP 10 8 EFPY m e.e LEAK TEST LIMIT _
[
/
/
L
)
(
(
~
s
\\
' l J
UNACCEPTABLE
[
/
/
OPERATIO:1
- m e.e f
r r
2
/
/
E i
(
HEATUPRATESUP
[
l
=
E TO 607/HR q
/
f
- y j
/
/
I
/
/
ACCEPTABLE e
g
/
/
OPERATION 2
(
m, J
/
,/
-- CRITICALITY LIMIT BASED ON INSERVICE
/
HYDROSTATIC TEST TEMPERA 1URE (274'F)
FOR THE SERVICE PERIOD y
UP TO 3 EFPY l
r e.e -
e.e toe.e see.e m.e see.e see.e ISOICATED TERPERATURE (840.F)
FIGURE 5 FARLEY UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 8 EFPY ______
MATERIAL PROPERTY BASIS CONTROLLI!G MATERIAL :
R. V. INTERMEDIATE SHELL COPPER CONTENT
- 0.20 WT%
PHOSPHORUS CONTENT
- 0.018 WT%
INITIAL RT
- 3/4T, 83 F CURVES APPLICABLE FOR C00U)0WN RATES UP TO 100 F/HR FOR HE SERVICE PERIOD UP 'It) 8 EPY n00.0
)
/
r l
3 ant.0 UNACCEPTABLE OPERATION
)
E l
ens:
(
j ACCEPTABLE g
OPERATION e.
)
a
{
E
/
5
/
/
1994.0 A[
TA'**'
O
~ E0h ::E:BS7
~
40 " W /
60 ' /
_ 100 '
0.0 0.0 100.0 400.0 000.0 400.0 900.0 I
}
IN0!CATED TERPERATURE (CEG.F)
FIGURE 6 FARLEY UNIT 2 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE FOR THE FIRST 8 EFPY l !
,