ML20084C100
| ML20084C100 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/20/1984 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20084C099 | List: |
| References | |
| TAC-54449, TAC-54894, NUDOCS 8404270083 | |
| Download: ML20084C100 (7) | |
Text
.
S Attachment Unit 1 Page 3/4 4-28 Page 3/4 4-29 Page 3/4 4-30 Page B3/4 4-7 Page B3/4 4-8 Page B3/4 4-9
- O l
l 4
P..
TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE VESSEL LEAD WITHDRAWAL CAPSULE LOCATION FACTOR TIME Y
343' 3.12 Removed 1.13 EFPY U
107' 3.12 Removed 3.02 EFPY X
287*
3.12 6 EFPY W
110' 2.70 12 EFPY V
290*
2.70 21 EFPY Z
340' 2.70 Standby FARLEY - UNIT 1 3/4 4-28 AMENDMENT NO.
M ATERIAL PROPERTY BASIS CONTROLLING MATERIAL: WELD METAL COPPER CONTENT: 0.24 WT%
PHOSPHORUS CONTENT: 0.011 WT%
0 RTNDT INITIAL: 0 F RTNDT AFTER 7 EFPY: 1/47,1850F 3/4T, 87'F CURVE APPLICABLE FOR HEATUP RATES 0
UP TO 60 F/HR FOR THE SERVICE PERIOD UP TO 7 EFPY AND CONTAINS
. ACCEPTABLE 0
3000 MARGINS OF 10 F AND 60 PSIG FOR REGION FOR POSSIBLE INSTRUMENT ERRORS HYDROSTATIC TESTING OPERATIONS 2y LEAK TEST LIMIT E
E 2000 UNACCEPTABLE a.
OPERATION S'
2o5 HEATUP RATES 3 ACCEPTABLE E
UP TO 60 F/HR OPERATION 0
p CRITICALITY LIMIT F BASED ON INSERVICE HYDROSTATIC TEST TEMPERATURE 0
(325 F) FOR THE -
SERVICE PERIOD UP TO 7 EFPY 0
0 100 200 300 400 500 INDICATED TEMPERATURE ('F)
Figure 3.4-2 Farley Unit I Reactor Coolant System Heatup Limitations Applicable For The First 7 EFPY FARLEY-UNIT 1 3/4 4-29 AMENOMENT NO.
MATERIAL PROPERTY GASIS CONTROLLING MATERIAL: WELD METAL COPPER CONTENT: 0.24 WT%
PHOSPHORUS CONTENT: 0.011 WT%
RT 0
NDT INITIAL:0 F RTNDT AFTER.7 EFPY: 1/4T,1850F.
3/4T, 87'F CURVE APPLICABLE FOR COOLDOWN RATES UP TO 100 F/HR FOR THE SERVICE '
0 PERIOD UP TO' 7 EFPY AND CONTAINS MARGlNS OF 100F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS
_ 2000 0
Di a:.
ED N
LINACCEPTABLE '
g OPERATION a-f S
ti
!:2
'_E 1000 ACCEPTABLE OPERATION COOLDOWN RATES OF/HR '
on
)
j 40'y 20 -
s W
60 100 0
O 100' 200 300 400 INDICATED TEMPERATURE (OF)
Figure 3.4-3Farley Unit 1 Reactor Coolant System Cooldown Limitations Applicable For The First 7 EFPY FARLEY-UNIT 1-3/4 4-30 AMENDMENT NO.
/
J REACTOR COOLANT SYSTEM BASES
- 4) The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200*F/hr respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F.
- 5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.
The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with ASTM E185-82 and in j
accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."
Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTndt, at the end of 7 effective full power years of service life. The 7 EFPY service life period is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unirradiated material. The selection of such a limiting RTndt assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirenents.
The reactor vessel materials have been tested to determine their initial RTndt; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTndt. Therefore, an adjusted reference temperature, based upon the Muence and copper content of the material in Regulatory Guide 1.99, Revision 1, gure B 3/4.4-1 ar.d the recommendations question, can be predicted using Fi Effects of Residual Elements on Predicated Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTndt at the end of 7 EFPY (as well as adjustments for possible errors in-the pressure and temperature sensing instruments).
i I
_ FARLEY - UNIT l' B 3/4 4-7 AMENDMENT NO.
,-h-r m-
-~
w e
, -, ~,.
d REACTOR COOLANT SYSTEM BASES Values of ARTndt determined in this manner may be used until the results from the material surveillance program, evaluated according to ASTM E185-82, are available. Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10CFR50, Appendix H.
The surveillance specimen withdrawal schedule is shown in Table 4.4-5.
The heatup and cooldown curves must be recalculated when the6*RTndt determined from the surveillance capsule exceeds the calculated ARTndt for the equivalent capsule radiation exposure.
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10CFR Part 50 and these methods are discussed in detail in WCAP-7924-A.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.
In the calculation procedures a semi-elliptical 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 G of ASME Section III 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 provide sufficient safety margins for protection against non-ductile 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, RTndt, is used and this includes the l
radiation induced shift, ARTndt, corresponding to the end of the period for which heatup and cooldown curves are generated.
I l
FARLEY - UNIT 1 B 3/4 4-8 AMENDMENT NO.
t
TABLE B 3/4.4-1
~.. ',
REACTOR VESSEL TOUGHNESS DATA y
Material Cu P
T M
RT Upper Shelf Energy NDT NDT y
Component Code No.
Type
(%)
(%)
_(*FJ
,(*F1 Q
MWD NMWD
)
Z Closure head done B6901 A533B, C1.1 0.16 0.009
-30 20 40
-20 140 Closure head segment B6902-1 A533B, C1.1 0.17 0.007
-20
-10 10(a)
-20 138-Closure head flange B6915.1 A508, C1.2 0.10 0.012 60(a)
--20 0(a) 60 75(a)
Vessel flange B6913-1 A508, C1.2 0.17 0.011 60(a) -30
-10(a) 60 106(a)
Inlet nozzle B6917-1 A508, C1.2 0.010 60(a) 45 60 110 Inlet nozzle B6917-2 A508, C1.2 0.008 60(a) 115 60 80 Inlet nozzle B6917-3 A508, C1.2 0.008 60(a) 35 60 98 Outlet nozzle B6916-1 A508, Cl.2 0.007 60(a) 60 60 96.5 0.011 60(a) 30 60 97.5 Outlet nozzle B6916-2 A508, C1.2 0.009 60(a) 50 60 ea Outlet nozzle B6916-3 A508, C1.2 100 0.010 30 70 90(a) 30 148 Nozzle shell B6914-1 A508, C1.2
'4 Inter. shell B6903-2 A533B, C1.1 0.13 0.011 0
-25 40 0
151.5 97 Inter. shell B6903-3 A533B, C1.1 0.12 0.014 10 5
52 10 134.5 100 Lower shell B6919-1 A533B, C1.1 0.14 0.015
-20
-5 75 15 133 90.5 Lower shell B6919-2 A533B, Cl.1 0.14 0.015
-10 0
65 5
134 97 0.010 10
-25
-5(a) 10 163.5 Bottom head ring B6912-1 A508, C1.1 g
Bottom head segment B6906-1 A533B., C1.1 0.15 0.011
-30
-50
-30(a)
-30 147 E
Bottom head done B6907-1 A533B, C1.1 0.17 0.014
-30
-10
-10(a)
-30 143.5 Inter. shell long.
0.27 0.015 0(a)
<60 0
-4 weld seams f
Inter. 10 lower shell 0.24 0.011 0(a)
<60 0
weld seam Lower shell'long.
0.17 0.022 0(a)
<60 0
weld seam
(*) Estimated per NRC Regulatory Standard Review Plan, section 5.3.2.
MWD - Major Working Direction NMWD - Normal to Major Working Direction f