ML20073K379
| ML20073K379 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/03/1994 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20073K377 | List: |
| References | |
| NUDOCS 9410110196 | |
| Download: ML20073K379 (31) | |
Text
s g
e ENCLOSURE 3 VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REVISION TO REACTOR PRESSURE LIMITS MARKED UP TECHNICAL SPECIFICATION PAGES 9410110196 941003 PDR ADOCK 05000424 P
l i
l REPLACE u) TH rJ ew Pa*e 3/4 of 5l
- 3000, URVE APPLICABLE FOR THE SERVICE RIOD UP TO 13 EFPY RTN
. After 13 EFPY g
7 a.1 T < 110* F S
- b. 3/ T< 87'F LEAK TEST LIMIT g
CRITICALITY LIMIT y
FOR 60*F/hr n.
2000 UNACCEP BLE HEATUP
[
OPERATI CRITICALITY m
j LIMIT jf n
FOR 100'F/hr HEATUP g
o m
60*F/hr o
HEATUP BASED ON INSERVICE '
g 1000 CURVE,f/
HYDROSTATIC TEST e
TEMPERATURE (255'F) g g
FOR THE SERVICE PERIOD UP TO 13 EFPY h
100'F/hr kHEATUP ACCEPTABLE CURVE OPERATION I
0.0 0.0 100 200 300 400 500 LOWEST INDICATED RCS Tcold TEMPERA RE ('F)
MATERIAL BASIS Copper Content:
0.10 Wt %
(
c.e4 wt %)
I
.40'F NDT ntttet:
RT (Actuel.
'F)
After 13 EFPY e 1/4 110'F RTNDT e 3/4 7 7'F FIGURE 3.4-2a UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 13 c PY V0GTLE UNITS - 1 & 2 3/4 4-31 Amendment No. 36 ( nit 1)
Amendment No. 16 (U it 2)
2500 g
g LeakTest Limit '
1 N
I,I 2250 2000 1750 l
Unacceptable l
Operation
]
3 1500 r
ta l
S E
a 1250 E
Heatup Rate up to N Acceptable n.
60 F/hr Operation 1000 j
s Heatup Rate up tos
.E 100 F/hr N
750 f
Criticality,Umit Based
)
/
on Insenace Hydrostatic
/
Test Temperature 500 yp (246 F) for the Service Period up to 16 EFPY s
250 i
OL l
0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature ('F)
MATERIAL BASIS l
Copper Content:
Assumed - NAWT%
(Actual- 0.083 WT%)
NDT nitial:
Assumed - NA *F I
RT i
(Actual-20'F)
I RTNDT At 16 EFPY: @ 1/4T - 100.7*F
@ 3/4T - 84.1*F Figure 3.4-2a l
Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates up to 100'F/hr) Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and Reactor Vessel Beltline Region).
i f
VOGTLE UNITS 1 & 2 3/4 4-31 l
l l t REPcue u>irw Neno PM e- %
'l-3l o
~
l/'
UNIT 2
/
i CURVE APPLICA8LE FOR THE
/,
SERVICE PERIOD UP TO 16 EFPY
/
l LEAK TEST l
y LIMIT S 2000 f
f f
f, f CRIT bALITY LIMIT FOR 3
0 60 F/hr HEATUP w
U EPTABLE E
OPERATION h
HEATUP CURVE
[ CRITICALITY LIMIT FOR e
0 100 F/hr HEATUP 8
ACCEPTABLE
)
OPERATION u
/
m 1000 i
l 100 F/hr HEATUP CURVE 0
E m
l 8
BASED ON INSERVICE l
O HYDROSTATIC TEST E
TEMPERATUR E (2880F) l 2 500 FOR THE SERVICE PERIOD UP TO 16 EFPY l
t 0.0 100 200 300 400 500 EST INDICATED RCS Tg TEMPER RE(OF)
MATEml A A$t$
I l
Cooper ce Asmened. 0.10 We %
( Astwas. 0.05 Wt %I I
NOT ""#
g r,,, ~ i..
.u.v.,a, j
e 3/. T = S P,
I l
FIGURE 3.4-2b UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 PY
\\
V0GTLE UNITS - 1 & 2 3/4 4-31a
2500 f
I Leak Test Limit J
fJ f
2250 l
l 2000 Unacceptable 1750 1
l g; 1500 l
Heat Rate Acceptable up to 0 F/hr Operaton l
1250 e
ec.uS 1000 B
,f
-3 Heatup Rate s
up to 100 F/hr 750 f
1 Criticality Umit Based 1
/ on Inservice Hydrostatic l
/
TestTem rature 1
500 1
- (256 F) f r the Service Period up to 16 EFPY j
s
\\
250 l
0 j
0 50 100 150 200 250- 300 350 400 450 500 Indicated Temperature ( F)
MATERIAL BASIS Copper Content:
Assumed - NA WT%
(Actual- 0.05 WT%)
NDT nitial:
Assumed NA'F RT I
(Actual-50*F)
At 16 EFPY: @ 1/4T - 112'F RTNDT
@ 3/4T = 94*F Figure 3,4-2b l
Unit 2 Reactor Coolant System Heatup Limitations (Heatup rates up to 100 F/hr) Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and Reactor Vessel Beltline Region).
VOGTLE UNITS 1 & 2 3/4 4-31a
l Rehe wem uew pus J/+ 4 31 UNIT 1 7
I 3000 g
j f
CURVE APPLICABLE FOR THE SERVICE i
l PERIOD UP,TO 13 EFPY l
3 O
E o
E E 2000 i
n.
1 3
i W
m UNACC TABLE
/
l OPERA ON ACCEPTABLE i
OPERATION j
)
5 O
i O
o e
C E 1000 2
5 COOLDOWN j
e o
RATE
]
('F/hr) 4
{
u Op 0
ff 60 /
100
\\
0.0 0.0 100 200 300 400 500 LOWEST INDICATED RCS Tcold TEMPE URE ('F)
MATERIAL BASIS copper content.
. o.to wt %
0.06 wt %)
NOT nfeet:
40*F
^ - ' -
IIT I
(Actuel.
- F)
After 13 EPPY e 1/4 T 110*F RTET e 3/4 T a "F
FIGURE 3.4-3a UNIT 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 13 E I YOGTLE UNITS - 1 & 2 3/4 4-32 Amendment No. 36 (Unit 1)
Amendment No.16 (Unit 2)
2500 l
2250 2000 l
Unacceptable Operation 1750
- 1500 b
e a
1250 e
Acceptable 0-Operation E
1000
.h
_E Cooldown Rates 'F/hr 500
- 8 9]V a0 250 RU" 0
0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature ( F)
MATERIAL BASIS Copper Content:
Assumed - NA WT%
(Actual- 0.083 WT%)
NDT nitial:
Assumed - NA 'F I
RT (Actual-20*F)
RTNDTAt 16 EFPY: @ 1/4T - 100.7'F
,t
@ 3/4T - 84,1*F Figure 3.4-3a Unit 1 Reactcr Coolant System Cooldown Limitations (Cooldown rates up to 100*F/hr)
Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and Reactor Vessel Beltline Region).
VoGTLE UNITS 1 & 2 3/4 4-32 l
5 l
kE@Lhce va t nt New PAae 3/+
y-32v 1,
l UNIT 2 i
i 1
CURVE APPUCABLE FOR THE SERVICE PERK)O UP TO 16 EFPY t
a 2000
(
w i
E l
lE w
k M
l t
m0 a
z UNACCE LE OPERATI g
4 1
o E
1000 g
W i
E ACCE 'TANLE J
COOLDOWN OPERATK)N i
o w
R ATE (*F/hr) 1 e
/
]
j 0
o M0 20 -
g _-
Z to 100
\\
0.0 0.0 6
200 300
- 0 W
LOWEST H409CATED RCS Tg TEMPERA RE (OF)
MATERI AL BA Copper Content; Assurned 0.10 Wt %
4
( Actusi 0.05 Wt %)
'4,7 = 97'F p
a FIGURE 3.4-3b UNIT 2 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EF a
i a
VOSTLE UNITS - 1 & 2 3/4 4-323
i 2500 2250 i
2000 l
f 1750
^
Unacceptable Operation g; 1500 ToS 2
a 1250 Acceptable Operation oS 1000 8e l
750 i
Cooldown l
Rates F/hr l
500
^
l 2!. 0 j.gg /
l 0
0 50 100 150 200 250' 300 350 400 450 500 Indicated Temperature ( F)
MATERIAL BASIS Copper Content:
Assumed - NA WT%
(Actual- 0.05 WT%)
NDT nitial:
Assumed NA'F I
RT (Actual-50'F)
RTNDTAt 16 EFPY: @ 1/4T - 112'F
@ 3/4T - 94'F i
Figure 3.4-3b Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100'F/hr)
Applicable for the First 16 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors and Margin of 74 psig for Pressure Difference Between Pressure Instrumentation and Reactor Vessel Beltline Region).
VOGTLE UNITS 1 & 2 3/4 4-32a i
l.
J i
3L:iLAC6 w im usw hA-6 e' J/+ 4-3 7 l
UNIT 1 M
N
/
l 1
-\\
/
E
-N
/
\\
(242,726)
/
(350,726) y
-\\
i
=
N I
/
O 700 N
I
/
i N
/
/
i N
//
I N
/
e
\\ //
2 000
- s.,-
l j
/\\/
^
l
/ A i
1 F
\\
,532)
-(70 g
1 if s
/f
\\f t
f f
f f f f
1 1
1 f
f f f f i f I
t I
i !
j 50 100 150 200 250 300 360 TRTD-A IONEERED LOW MEASURED RCS TEMPER TURE (*F) d 1
1 j
FIGURE 3.4-4a 4
UNIT 1 MAXIMUM ALLOWABLE NOMINAL PORY SETPOINT i
FOR THE COLD OVERPRESSURE PROTECTION SYSTEM
0GTLE UNITS - 1 & 2 3/4 4-35
800 780 760 740
-y (212,726)
(350 726)
S 720 E
J 700 o*
l 680 C
660 l
og 640 f
3:
- S 620 600
'Ey 580
$560
/
( 0,538) j m
540 i
520 500 50 100 150 200 250 300 350 TRTD - Auctioneered Low Measured RCS Temperature ( F)
Figure 3.4-4a Unit 1 Maximum Allowable Nominal PORV Setpoint for the Cold Overpressure Protection System VOGTLE UNITS 1 & 2 3/4 4-35 l
REPt/rce w en+
v e w 8kG c' 3/4 4-3 fo UNIT 2 800
\\
-\\
/
N
/
'N (2a2,725)
/
(350,728) f
{
~
N
/
~ 700 s
7 u
N
/
'\\
I/
~
/
x 2
\\
['
s N
f e
n
.00
\\
/
</
I
/ A i
/
\\
p (70,505)
[
\\
n.
g
\\
/
\\
7
/
\\
50 100 150 200 250
\\@B0 50 TRTD7'AUCTIONEERED LOW MEASURED RCS TEMPERA (*F)
FIGURE 3.4-4b UNIT 2 MAXIMUM ALLOWABLE NOMINAL PORV SETPOINT FOR THE COLD OVERPRESSURE PROTECTION SYSTEM V TLE UNITS - 1 & 2 3/4 4-35a
5 800 780 760 740 (222,726)
(350,726)
S 720
=
l 3, 700 j
o*
680 l
cco 660
/
I l
f 640
[
SR 620
]
600
/
1:y 580 h560
/
I 1
w 540
/
/
520 (70,516) 500 50 100 150 200 250 300 350 i
TRTO - Auctioneered Low Measured RCS Temperature ( F) l l
l l
Figure 3.4-4b Unit 2 Maximum Allowable Nominal PORV Setpoint for the Cold Overpressure Protection System VOGTLE UNITS 1 & 2 3/4 4-35a
}.
l BASE 5 PRES 5URE/ TEMPERATURE LIMITS (Continued) 2
~
2.
These limit lines shall be calculated periodically using methods provided
- below, 3.
T.he secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F, 4.
The pressurizer heatup and cooldown rates shall not exceed 100*F/h and 200*F/h, respectively.
The auxiliary spray shall not be used if the temperature difference between the pressurizer and the auxiliary spray fluid is greater than 625*F, and 5.
System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler L
\\
and Pressure Vessel Code,Section XI.
M 4A d q~Q The fracture toughness properties of the ferritic materials in the reactor Z
vessel are determined in accordance with the NRC Standard Review Plan, ASTM M
E185-82, and in accordance with additional reactor vessel reouirements. l Th;;;
- r
- ::rti::
" ::::rd:n:: with A;;:ndi C Of th: 1972 E r:r Add;nd; t: 5;;;i;r !!! ;f th; ASMC ";il;r ;nd "r;;;;re V;;;;l C;C; :ndi tne calculation metnods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," April 1975.
The heatup and co own limit curves.sho in Figures 3.4-2a an 3.4-3a are licable to Unit 1 up to 13 EFPY and a based on Westinghous l
/
develop generic curves whi were develo ed assum a 40*F initial RT and a copp content of 0.10 or the most limiting terial.
These h ves nh are applicabi o Unit i since its t limiting material ble B 3/4.4-la) 2 has both a lower ' itial RTNDT (30*F) a lower copper cent (0.06 WT%).
i e curves, however, re not applicable t nit 2, since its mos limiting mate 1 (Table B 3/4.4-has a higher initi RT (50 compared t 40*F).
arate heatup and idown limit curves yNdevelopedbasedo he actual mate al properties of t most limiting mater 1 for Unit 2 up to 1 EFPY, The Un curves are shown Figures 3.4-2b and 4-3b.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNOT, at the end of the Effective Full Power Years (EFPY) of service life. The EFPY service life period is chosen such that the limiting RT at the 1/4T location in the core region NDT is greater than the RT f the limiting unirradiated material.
The selection NOT of such a limiting RT assures that all components in the Reactor Coolant NDT System will be operated conservatively in accordance with applicable Code requirements.
V0GTLE UNITS - 1 & 2 B 3/4 4-8 Amendment No. 36 (Unit 1)
Amenament No.16 (Unit 2)
INSERT A These properties are then evaluated in accordance with Appendix G of ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Appendices, " Rules for l
Construction ofNuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure," 1986 Edition and I
INSERT B The heatup and cooldown limit curves shown in Figures 3.4-2a and 3.4-3a for Unit I and Figures 3.4-2b and 3.4-3b for Unit 2 are applicable for up to 16 EFPY and were developed based on the actual material properties of the most limiting material. The l
most limiting material are shown in Table B 3/4.4-la for Unit I and Tables B 3/4.4-lb i
for Unit 2.
I i
i I
i t
L l-1 l'
I
=
7
s
.I AB_L 4.4-la ONil 1 RI ACIOR _VI'$5t L_ IOUGHNL SS ASML
$0 I l-ll!
AVI'RAGL c
RI 5
COMP AL CU Ni P
N 35 MIL Nul HMWD*
d C
PJ CODE TYPE
(%)
1*B Eg ( I )
(*r).
(fI-tII)
Closure Hea B8807-1 A533BCL.l
.1
.67
.008 -50 15 15 88 l
e-Closure Ilead Tor B8808-1 A5330Cl.1
. 14
.010 -30 68 85 Closure Head Flange 8801-1 A508Cl.2
.70 11 20 (40 20 1 12
~
Vessel F lange B
-1 A508CL.2
.71
.01 0
<60 0
11 Inlet Nozzle B8809-A508CL.2
.86
.011
<10
-20 10/
Inlet Nozzle 88809-2 8CL.2
.84
.014
-10
<50
-10 95 let Nozzle B8809-3 A50
.82
.013 -10
-10 11/
Inl ozzle
-B8809-4 A508CL.2
.87 014
-20
<10
-20 105 Outlet le 88810-1 A508CL.2
.82
.006 -10
<50
-i
>124 Outlet Nozz 88810-2 A508CL.2 79
.006
-10
<50
-10 100 i
Outlet Nozzle 88810-3 A508Cl.2
.006 -10 (50
-10 R
Outlet Nozzle 8810-4 A508CL.2
.80
-10
<10
-10
/5 Nozzle Shell B
-1 A5338CL.1
.14
.62
.011 0
88 28 94 i
Nozzle Shell 88804-A533BCL.1
.10
.58
.006 -40 75 15 104 zzle Shell 88804-3 38CL.1
.14
.69
.013 -30 0
40 92 In Shell B8805-1 A53 1
.08
.59
.004 0
6 0
90 Inter.
11 88805-2 A533BCL.
.08
.59
.004 -10 80 40 100 Inter. Shel 88805-3 A533BCL.1
.60
.003 -20 90 36' 10/
i Lower SheiI B8606-1 A5338CL.1
.05 59
.005 -50 80 20
.116 Lower Shell 8606-2 A533BCL.1
.05
.009 -10 80 20 113\\
Lower Ehell B
-3 A533BCL.1
.06
.64 7 -20 10 10 118 Bottom Had Torus 88813-A5338CL.1
.13
.50
.009
-40 50
-10 88 ottom Head 00me B8812-1 3BCL.1
.10
.53
.009
-4 32
-28 122 In
& Lower Shell Gl.43 SAW
.03
.10
.00/
-80
-20
-80
- 129 Verti Weld Seams an th Seam.
- 0pper ShelI energy, norma i major working directions T
l i.
s l
' 15 D h,
i 11 k'
)/
.._-.._._._.._..~....__...._..._._____._.._.m.____.
m.__..._.____._..__..
i L
TABLE B 3/4.4-1a UNIT 1 REACTOR VESSEL TOUGHNESS g
j nm COMP CU NI INITIAL 16 EFPY RTNDT lil COMPONENT CODE 1%1 L%1 B L ui El 1/4-t ( F) 3/4-t i El
]
Closure Head Flange 0.70 20 Vessel Flange 0.71' O
?
Intermediate Shell 88805-1 0.083 0.597 0
80.7 64.1 r
intermediate Shell*
B8805-2 0.083 0.610 20 100.7 84.1 l
Intermediate Shell B8805-3 0.062 0.598 30 97.5 76.4 l
Lower Shell B8606-1
'O.053 0.593 20 77.6 57.6 i
Lower Shell B8606-2 0.057 0.600 20' 81.9 62.5 Lower Shell B8606-3 0.067 0.623 10 80.8 60.6 Circ. Weld 101-171 0.039 0.102
-80
-21.7
-39.9 l
Long. Weld 101-124A 0.039 0.102
-80
-31.8
-48.5 Long. Welt 101-124B 0.039 0.102
.-80
-30.0
-47.0 Long. W 101-124C 0.039 0.102
-80
-30.0
-47.0 3
Long. Weio 101-142A 0.039 0.102
-80
-30.0
-47.0 t
Long. Weld 101-1428 0.039 0.102
-80
-31.8
-48.5 Long. Weld 101-142C 0.039 0.102
-80
-30.0
-47.0 i
- Limiting material l
l I
i lt I
i-I 4
x IABLE B 3/4.4-Ib N
'\\
x x
o
'N uMIT 2 REACIOR VESSEL 100GHNE55 N
S N
ss g
s ASME AVI RAGE c-I NI
?!
COMP NAIERIAL C
Ni P
NDI N01 HMWI)"
d C3MPONENT CODE lYPE
{%}
{%)
(*f)
( *1' )
('r l-L8 )
[
Closure Head Gone R9-1 A5338 CL. 1 0.07 0.61 0.008
-40
-30 123 Closure Head Torus 0-1 A5338 CL. 1 0.07 0.64 0,010
-30 0
114 0.72 0.011.
10 to l io Closure Head Flange R7-A508 CL. 2 Vessel Flange Rl-1 A508 CL. 2 0.81 0.011 \\ 0
-60 115 m
6 89806-1 08 CL. 2 0.07 0.84 0.010
-50s.
-50 119
\\. Inlet Nozzle Inlet Nozzle 89806-2 A568' L. 2 0.06 0.83 0.009
-40
N,-40 128
~
thlet Nozzle RS-1 A508 CL.
0.09 0.87 0.008
-20
-20 141 InletNezzle RS-2 A508 CL. 2 0.08 0.85 0.009
-20
-20
. 134 Outlet Nott e R6-3 A508 CL. 2 0.69 0.011
-10
-10 122 Outlet Nozzle R6-4 A508 CL. 2 0.66 0.010
-10
-10 t40 Outlet Nozzle 89807-3 A508 CL. 2 0.005
-30
-30 116 R
Outlet Mozzle 9807-4 A508 CL. 2 0.64 0.010 10 10
- 132 Nozzle Shell R3-1 A5338 CL. 1 0.20 0.67 15 0
20 19
?
Nozzle Shell R3-2 \\ A5338 CL. 1 0.20 0.67 0.0l3 0
40 19 y
Nozzle Shell R3-3
'45338 CL. 1 0.15 0.62 0.010 IQ 60 84 CL. 1 0.06 0.64 0.009
-20 K 10 95 A$33(8C-
' Intermediate She1I R4-1 A533 1
0.05 0.62 0.009
-10 x10 104 Intermediate Shell R4-2 Inteniihilate Shell R4-3 A5338 CL. x 0.05 0.59 0.009 0
30N 84 Lower Sheh 88825-1 A5338 CL. 1 D 45 0.59 0.006
-20 40
s 83 Lower Shell R8-1 A5338 CL. 1
- 0. g o.62 0.007
-20 40
'B/s '
Lower Shell**
R8628-1 A5338 CL. 1 0.05 8 49 0.001
-20 50 85 s
0.012
-20
-20 89 H-Botton Head Torus 4tl2-1 A5338 CL. 1 0.17 0.64N '8.008
-30
-30 115 1
2)
Bottom Head Dome Rll' A5338 CL. 1 0.10 0.62 n
Shell Vertical
([g i oy ', /
Intermediate & Lower Gl.60 SAW 0.07 0.13 0.007s -10
-10 141
\\
L l(Seams Inte
' ate to Lower E3.23 SAW 0.06 0.12 0.001
-50
-30 90
(/
Shell Gt Weld 3
Seam
" Upper Shelf energy,Dr al to major working direction.
- Limiting material.
TABLE B 3/4.4-1H t
UNIT 2 REACTOR VESSEL TOUGilNESS COMP CU NI INITIAL 16 EFPY RTNDT COMPONENT CODE RTndt 1/4-t (*F) 3/4-t(*F)
LD
(
Closure IIcad Flange 0.72 10 Vessel Flange 0.87
-60 Intermediate ShcIl R4-1 0.06 0.64 10 81 62 Intermediate Shell R4-2 0.05 0.62 10 72 54 Intermediate Shell R4-3 0.05 0.59 30 92 74 Lower Sell B8825-1 0.05 0.59 40 102 84 f-1 i
Lower Shell R8-1 0.06 0.62 40 111 92 L
t n
Lower Shell*
B8628-1 0.05 0.59 50 112 94 P
-l Cire. Weld 0.06 0.12
-30 55 31 O
Long. Weld 0.07 0.13
-10 83 56 i
- Limiting Material 4
4 l
j i
l BASES i
PRESSURE / TEMPERATURE LIMITS (Continued)
}
The reactor vessel materials have been tested to determine their initial RT j
NOT: the results of these tests are shown for Units 1 and 2 in Table
~
1 B 3/4.4-la and b, respectively.
Reactor operation and resultant fast neutron i
J (E greater than 1 MeV)-irradiation can cause an increase in the RT There-l d
ore,anadjustedreferencetemperature,baseduponthefluence,cohe.r content, y
=
C an : n ::n r e content of the material in question, can be predicted using lFi;;r: 5 2/'
"-1 :nd th: ur;::t :: :: :" ^ % : r;;t:c by ::tn:q Regulatory Guide 1.99, Revision 2, " Effects of Residual Elements on Predicted Radiation l
Damage to Reactor Vessel Materials," Ler the Weetinenesee Cesser Trenc Curve;l i
1 :n:r " "i;_r; O 2/4.4 2!
The heatup and cooldown limit curves of Figures 3.4-2a and 3.4-3a (Unit 1), Figures 3.4-2b and 3.4-3b (Unit 2) include predicted l
f (o adjustments for this shift in RT at the end of il; (Jan l' :n: x (Unt
- n l
NOT l
EFPY as well as adjustments for possible errors in the pressure and temperature I
sensinginstrumentsj p 60 p.3 and lov, ruput'gejy
~
j l
Values of MT determined in this manner may be used until the results NOT j
from the material surveillance program, evaluated according to ASTM E185~ are available.
Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR Part 50, Appendix H.
The surveillance specimen with-i drawal schedule is shown in Table 16.3-3 of the VEGP FSAR.
The lead factor i
represents the relationship between the fast neutron flux density at the loca-tion of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future y,
radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule.
The heatup and cooldown curves must be i
recalculated when the MT determined from the surveillance capsule exceeds i
NOT the calculated M T for the equivalent capsule radiation exposure.
NOT Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following.
paragraphs.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) tech-nology.
In the calculation procedures a semielliptical surf ace 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 capaoilities of inservice inspection techniques.
Therefore, the reactor operation limit cur-ves developed for this reference crack are conservative and provide sufficient
~
safety margins for protection against nonductile failure.
To assure that the radiation emorittlement effects are accounted for in the calculation of the V0GTLE UNITS - 1 & 2 B 3/4 4-10 Amendment No. 36 (Unit 1)
Amenoment No.16 (Unit 2)
INSERT E In addition, these curves include a pressure adjustment of 74 psig to account for the pressure difTerential between the wide range pressure transmitter and the belt line region.
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' BASES 1
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PRESSURE / TEMPERATURE LIMITS (Continued) be analyzed in order to assure that at any coolant temperature the lower value i
of the allowable pressure calculated for steady-state and finite heatup rates j
is obtained.
l The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T. deep outside surface flaw is assumed.
Unlike the situation at the vessel inside surface, 4
i the thermal gradients established at the outside surface during heatup produe.e l
stresses which are tensile in nature and thus tend to reinforce any pressure i
stresses present. These thermal stresses, of course, are dependent on both j
the rate of heatup and the time (or coolant temperature) along the heatup i
ramp. Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
i Rather, each heatup rate of interest must be analyzed on an individual basis.
i j
Following the generation of pressure-temperature curves for both the steady-j state and finite heatup rate situations, the final limit curves are prod ed as j
follows. A composite curve is constructed based on a point-by-point c rison of the steady-state and finite heatup rate data. At any given temperatusy, the l
allowable pressure is taken to be the lesser of the three values taken fWe the i
curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations 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.
Next, the 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.
l Finally, the new 10CFR50 Appendix G Rule which addresses the metal temperature
(
of the closure head flange and vessel flange regions is considered. This rule states that the minimum metal temperature of the closure flange regions should be
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th+1 j
at least.120*F higher than the limiting RT, for these regions when the pressure exceeds reent of the preservice hydrostatic test pressure (621 psig for l
For Unit I the minimum temperature of the closure flange
)
and v ions isJ40*F, since the limiting RT, is 20*F (see i
Table B 3/4-4.la).104 "q,tle %;t 1 teetei, cerve ste r. er. Fisure -t.2e is l
- W::t;d by th;.n.100F",0 rule.
";_.4;r, tra niitle %it I ceeld;, cerve i
t h r_
'n Ff;;r; 2 t. e is i;s,ectM to ite.... iGCFR3G ruiu. I For Unit 2, the j
minimum temperature of the closure flange and vessel flange regions is 130*F.
j since the limiting RTwis 10'F (Table B 3/4-lbM The Wit 2 5::typ ;re; j
s m " 9;re3.'-25 d th: :: ldr_; : r;: :hr_; in Fig;re 3.t-Ob er n;t j
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HER E V0GTLE UNITS - 1 & 2 B 3/4 4-15 i
INSERT F These values include margin of 10 F and 60 psig for instrumentation errors. The heatup and cooldown curves as shown in Figures 3-4.2a and 3-4.3a for Unit I and the heatup and cooldown curves as shown in figures 3-4.2b and 3-4.3b for Unit 2 are impacted by the new 10 CFR 50 rule.
1 l * "
BASES.
i i
PRESSURE / TEMPERATURE LIMITS (Continued) l Although the pressurizer operates in temperature rcnges above those for which there is reason for concern of nonductile failure, operating limits are provided to s
l assure compatibility of operation with the fatigue analysis performed in accordance j
with the ASME Code requirements.
I j
COLD OVERPRESSURE PROTECTION SYSTEMS The OPERABILITY of two PORVs, two RHR suction relief valves, a PORV and RHR SRV, or an RCS vent capable of relieving at least 670 gpa water flow at 470 psig ensures j
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 4
equal to 350*F. The PORVs have adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either:
(1) the start of an idle i
RCP with the secondary water temperature of the steam generator less than or equal to i
50*F above the RCS cold leg temperatures, or (2) the start of all three charging pumps I
and subsequent injection into a water-solid RCS.
The RHR SRVs have adequate relieving
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capability to protect the RCS from overpressurization when the transient is limited to either:
(1) the start of an idle RCP with the secondary to primary water temperature i
difference of the steam generator less than or equal to 25'F at an RCS tempera e of i
350'F and varies linearly to 50*F at an RCS temperature of 200*F or less, or (2 the i
start of all three charging pumps and subsequent injection into a water-solid A
j combination of a PORV and an RHR SRV also provides overpressure protection for the i
RCS.
j The Maximus Allowed PORV Setpoint for the Cold Overpressure Protection System
{
(COPS) is derived by analysis which models the performance of the COPS assuming j
various mass input and heat input transients. Operation with a PORV Setpoint less j
than or equal to the maximum Setpoint ensures that the nominal M = im unn 1.nm 16 EFPY = Udt 0 Appendix G reactor vessel NOT limits criteria will not be violated j
with consideration for a maximum possure overshoot beyond the PORV setpoint which can j
occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more i
severe than those assumed cannot occur, Technical Specifications require lockout of l
all safety injection pumps while in MODES 4, 5, and 6 with the reactor vessel head in-j stalled and disallow start of an RCP if secondary temperature is more than 50*F above i
primary temperature. Additional temperature limitations are placed on the starting of i
a Reactor Coolant Pump in Specification 3.4.1.3.
These limitations assure that the RHR system remains withia its ASME design limits when the RHR relief valves are used j
to prevent RCS overpressurization.
j The Maximum-Allowed PORY Setpoint for the COPS will be updated based on the results of examinettons of reactor vessel material irradiation surysOlance specimens i
performed as required 'oy 10 CFR Part 50, Appendix H, and in accordance with th j-i schedule in Table 16.3-3 of the VEGP FSAR.
4 3/4.4.10 STRUCTURAL INTEGRITY u
i The inservico inspection and testing programs for ASME Code Class 1, 2, and 3 j
components Nure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.
i j
These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addonda as required by 10 CFR 50.55a(g) except where i
i specific written relief has been granted by the Commission pursuant to 10 CFR j
50.55a(g)(6)(i).
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V0GTLE UNITS - 1 & 2 B 3/4 4-16 Amendment No. 56 (Unit 1) i Amendment No. 35 (Unit 2)
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i flGURE B 3/4.
EffECT LUENCE AND COPPER CONTENT'ON IFT OF RTNDT FOR REACT ESSELSEXPOSEDTORADIATIONA(5500F s
ENCLOSURE 4 VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REVISION TO IEACTOR PRESSURE VESSEL LIMITS EXEMPTION REOUEST This enclosure requests, in accordance with the provisions of 10 CFR 50.12, an exemption from certain requirements of 10 CFR 50.60, " Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation." As stated in 10 CFR 50.60, " Proposed Alternatives to the Described Requirements in Appendices G and H of this part or portions thereof may be used when an exemption is granted by the Commission under 50.12." This exemption is requested to allow the application of the l
American Society ofMechanical Engineers (ASME) Code Case N-514, " Low Temperature Overpressure Protection," in determining acceptable setpoints for VEGP Units 1 and 2.
Pressure / temperature (P/T) limits for low temperature overpressure protection are l
characterized by the system enabling temperature and the setpoint pressure for the pressure relieving device. Current regulatory guidelines require that the low temperature overpressure protection (LTOP) system must be enabled at temperatures less than or equal to R FNDT + 90 F, where RT i
NDT s the adjusted reference temperature including margin, at the quarter thickness location. At temperature greater than RTNDT + 90 F LTOP protection need not be provided. The maximum LTOP system pressure is determined based on system-specific considerations, but is chosen so that the maximum pressure attained in the vessel will not exceed the P/ f limit curve defined by Appendix G l
to ASME Section III and Appendix G to 10 CFR 50.
The LTOP limits caused operational constraints by limiting the range available to the operator to heat up and cool down the plant. The operating window through which the reactor coolant system is heated up and cooled down is determined by the difference i
l between the maximum allowable pressure determined from ASME Section III, Appendix G, and the minimum allowable pressure determined by the differential pressure between reactor coolant system pressure and atmospheric pressure for the RCP seals. As previously discussed, the Westinghouse methodology used to calculate LTOP setpoints did not account for the differential pressure across the reactor core during RCP operation.
E4-1 l
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i 1
e ENCLOSURE 4 (CONTINUED)
VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REVISION TO REACTOR PRESSURE VESSEL LIMITS EXEMPTION REOUEST l
The existing LTOP setpoint curve is based on a plant-specific evaluation with PORV setpoints selected within a range of allowable pressures at various temperatures. For the existing curves, the 74 psi nonconservatism depletes the existing margin at approximately 1200F for Unit I and 1450F for Unit 2.
1 ASME Code Case N-514 requires that the LTOP systems limit the maximun pressure in the vessel to 110 percent of the pressure determined to satisfy Appendix G limits.
The ASME Working Group on Operating Plant Criteria developed code guidelines to define LTOP limits that will avoid unnecessary operational restrictions, provide adequate margins against failure, and reduce the potential for unnecessary activation of pressure-relieving devices used for LTOP The LTOP limits allow the pressure that may occur with activation of pressure-relieving devices at temperature less than 2000F to exceed the P/T limits, however, acceptable margins to vessel fracture are maintained during these events.
The pressure vesselis protected both from LTOP events and the P/T limits in Technical Specifications applicable for normal heatup and cooldown in accordance with Appendix G i
to 10 CFR 50 and Sections III and XI of the ASME code.
Some conservatism in Appendix G pressure and/or temperature curve calculations are:
- 1. Safety factor of 2 on the principal membrane (pressure) stresses.
- 2. The disregarding ofincreased mechanical properties of the vessel that accompany material embrittlement (element yield strength and flow stress).
- 3. The limiting toughness is based upon a reference value (KIR), which is a lower bound of the dynamic crack initiation or arrest toughne s.
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ENCLOSURE 4 (CONTINUED)
VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REVISION TO REACTOR MESSURE VESSEL LIMITS EXEMPTION REOUEST
- 4. A margin factor of 2a (sigma) applied in determining the adjusted reference temperature (ART).
- 5. An assumed flaw in the wall of the reactor vessel with a depth equal to 1/4 of the thickness of the vessel wall and a length equal to 1 1/2 times the vessel wall thickness.
Bases for Exemotion l
The requested exemption to the regulations is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security.
Georgia Power Company believes that the requested exemption meets the criteria in 10 CFR 50.12 (a)(2)in that special circumstances are present. These includes:
The application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.
The basis for the LTOP setpoints is to preclude the potential for brittle failure of reactor vessel material. ACME Code Case N-514 recognizes the conservatism of the Appendix G curves and allows establishing a setpoint that preserves the acceptable margin of safety while maintaining operational rnargins for RCP operation at low temperatures and pressures. Setpoints established in accordance with Code Case N-514 will also minimize the unnecessary activation of protection system pressure relieving devices. Therefore, establishing LTOP setpoints using Code Case N-514 criteria satisfies the underlying purpose of the ASME code and regulations that nuclear power plant systems and components are operated to ensure an acceptable level of safety and environment impact.
E4-3
s.
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ENCLOSURE 4 (CONTINUED)
VOGTLE ELECTRIC GENERATING PLANT REQUEST TO REVISE TECHNICAL SPECIFICATIONS REVISION TO REACTOR PRESSURE VESSEL LIMITS l
EXEMPTION REOUEST Based on the above, the application of the regulation in the particular circumstances is not necessary to achieve the underlying purpose of the rule.
10 CFR 50.12 (1)(2)(iii)
Compliance would result in undue hardship or other costs that are significantly in l
excess of those contemplated when the regulation was adopted, or that are j
significantly in excess of those incurred by others similarly situated.
l Administrative restrictions on RCP operations while at low RCS temperatures would result in an unnecessary burden in that a long delay would be required to ensure minimum RCS temperature before starting the RCPs. The application of the code case will minimize any delays in starting the plant. The proposed LTOP P/T limits provide an acceptable margin against crack initiation and failure in reactor vessels. The P/T j
limits do not significantly change the likelihood of vessel failure associated with normal l
heatup and cooldown limits. The LTOP limits also reduce the potential for unnecessary activation of pressure-relieving devices. The new LTOP limits provide both economic and safety benefit.
Therefore, compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted.
Conclusion ASME Code Case N-514 allows setting the LTOP setpoints such that the Appendix G curves are not exceeded by more than 10 percent. At this time, GPC only wants to use this code case at temperature less than 2000F. The ASME code committee has concluded the LTOP guidelines provide acceptable margin against crack initiation and failure in reactor vessel and will reduce the potential for unnecessary activation of protection system pressure-relieving desdces. Consequently, the proposed LTOP setpoints provide both operational and safety benefits with no adverse safety or environmental impact. GPC believes that use of Code Case N-514 provides an acceptable level of quality and safety.
E4-4
- f. o l
l ENCLOSURE 4 (CONTINUED) j VOGTLE ELECTRIC GENERATING PLANT l
REQUEST TO REVISE TECHNICAL SPECIFICATIONS REVISION TO REACTOR PRESSURE VESSEL LIMITS EXEMPTION REOUEST l
Compliance with the current approved Pff limits would result in economic hardship to GPC without a compensating increase in he level of quality or safety. Georgia Power Company requests that this exemption from certain requirement of 10 CFR 50.60 be processed by Maren 31,1995.
l l
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E4-5 l
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