ML20199L937
| ML20199L937 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 11/25/1997 |
| From: | Tulon T COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9712020152 | |
| Download: ML20199L937 (17) | |
Text
Commonrecalth iden Company
!Wid jmx! 6< twrating Wiion
. Route el, Ika H 4 Hrurville 1160 tO7Axi]9 TelHlbl M 2Hul November 25,1997 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington. D. C. 20555 - 0001
Subject:
Reactor Vessel Shell Weld Indication Evaluation Braidwood Nuclear Power Station, Unit 2 Facility Operating License NPF-77 NRC Docket Number: 50-457
Reference:
(1)
T.J. Tulon (Commonwealth Edison) to US Nuclear Regulatory Commission letter dated October 15,1997.
(2)
T.J. Tulon (Commonwealth Edison) to US Nuclear Regulatory Commission letter dated October 16,1997.
(3)
G.F. Dick, Jr. (US Nuclear Regulatory Commission) to 1. Johnson (Commonwealth Edison) letter dated October 16,1997.
In Reference (1) Commonwealth Edison (Comed) provided the evaluation ' which demonstrated that an indication in the Braidwood Unit 2 Reactor Vessel, detected during the recent Unit 2 refueling outage, was acceptable for continued sersice. Reference (2) provided an aflidavit disclosing the proprietary nature of the evaluation and also provided the applic'ation for Aithholding the information from public disclosure.
Based on the review of the flaw evaluation handbook, NRC Staff requested Comed to verify one case involving a particular flaw size, shape, and orientation under a particular transient. The Comed response to this NRC request for additional information is contained in the attachment to this letter.
- WC AP 12045 and 12046, Handbook on Flaw Evaluation for Zion, Byron, and Braidwood Reactor g
Pressure Vessels"
@j 9712O20152 971125 8'#
llRElH]IlljllEMI
^"
A 1:niwm Comp.my
. ~ -
..~..
m,; ' ~
~
- i. S
.4:
>F,+
=
- V';
e
~'
' Document Control Desk November 25,1997 l
=
Page 2?
The choice o'r flaw and transient was discussed in a telecon conducted between 'NRC iStaff and Comed on October 23,1997.LThe requested 'infortnation related to the -
flaw / transient scenario was provided to Comed in Reference-(3) and modified in a telecon between Westinghouse and NRC on November 3,1997.
. Please address any comments or questions regaiding this matter to T.W. Simpkin at (815) 3 458-2801 extension 2980c Sincerely,
/
thy J. Tulon r
te Vice President-
- Braidwood Nuclear Generating Stat.on 4
Attachment:
" Summary of Flaw Evaluation for Braidwood Unit 2" 1 A.B'. Beach - Regional Administrator-Rill cc:
O.F. Dick, Jr. - Braidwood Project ht. nager-NRR C.J. Phillips - Senior Resident Inspector-Braidwood
- J.A. Gavula - Rill.
. Office ofNuclear Safety -IDNS;
.- o;. ntc\\97126mt doc j ua 3 '
- . p
, ~..,.. _.
SlStMARY OF FLAW EVALUATION FOR BRAIDWOOD UNTT 2 i
This note will briefly describe *he process used to develop the flaw evaluation chart used to disposition the indication recently discovered in the upper shell transition weld of Braidwood Unit 2. The indication was actually located near the outside surface of the reactor vessel, and was circumferential!y orientated. Although the indication was located far :nough from the surface to qualify as embedded, by the rules of Section XI, it was shown to be acceptable whether it was a surf ace or embedded indication. This example will be carried out to determine the largest allowable surface flaw, which was one of the results submitted.
For outside surface flaws in the upper shell transition region the most severe transients are the pressure transients, since the thermal transients result in tensile stresses near the insive surface, and compressive stresses in the region of interest here. There are no transients which result in a rapid heat-up, which would produce tensile thermal stresses in this region. The governing transients are:
Cold Hydrotest @ 3105 psi (normal / upset)
Large Steamline Break (emergency / faulted)
It should be mentioned that residual stresses are known to exist in this weld, but since the reactor vesselis stress relieved, the stress values are small, Measurements of stress relieved heavy section welds have shown residual stresses of about 5 ksi at each surface, with the stresses decreasing and becoming compressive in the center of the weld These stresses are present at a'l times, and will have an effect on fracture at low temperatures, when the toughness is Iri the transition region. At higher temperatures such as these in the region of interest here, the residual stresses have no effect on the failure conditions. This has been demonstrated experimentsily in the Heavy Section Steel Technology Intermediate Vessel test program. Therefore, residual stresses have not been used in the calculations discussed for this region, it will be seen from Figure A-2.5 of WCAP 12046 (reproduced here as Figure 1) that the allowable depth for any indication, regardless of shape, is at least 20 percent of the wall thickness. The allowable depth line is across the very upper edge of the figure.
This line is the result of 3 direct application of the Section XI acceptance criteria.
The allowable flaw depth is determined by calculation of the stress intensity factor (K) as a function of postulated flaw depth for each of the governing transients, and then determining where the K value exceeds the allowable toughness. We will follow this calculation in detail for the normal / upset case first. The stress distribution from a detailed finite element model of the reactor vessells plotted in Figure 2. The stress intensity factor was then calculated for three different flaw shapes, and these results are shown in Figure 3. The reference used for this calculation was Raju and Newman (reference 7 of WCAP 12045, rev.1).
l
(
0 %p antwntletters Thw Eval ror Draidwed 2
?.-
The allowable toughness was determined by reducing the fracture toughness by the factor V10, as required by Section XI. The hydrotest at 3105 is only conducted before operation, but for conservatism here it is assumed to occur during service. The hydrotest temperature for Braidwood Unit 2 has been calculated as 200 F for the leak test (WACP 14970) and will be higher for the 3105 hydrctest. At 200 F for the Braidwood 2 Nozzle Shell to Intermediate Shell Weld, the fracture toughness will be on the upper shelf, since there is no irradiation effect, and the initial RTuor for this weld is 25F. The allowable toughness is then 200/V10 = 63.2 ksi sq rt in, and this value is also plotted in Figure 3.
The following allowable flaw depths for normal / upset conditions result from these calculations:
Flaw Shape (a/l)
Allowab;e Depth (a/t) 0.01 0.269 0.1667 0.355 0.5 0.869 The flaw evaluation chart is then determined from the worst case of the results above and the results for the governing faulted condition. For the steamline break, the highest stress in the region of the flaw is early in the transient. The worst time step is at 100 secs., as may be seen from Table 1, and is plotted in Figure 4. The temperature, pressure and flow rate vs. time for this transient are provided in Appendix 1. Similar stress intensity factor calculations were done for this case, and the results were very low stress intensity factors, because of the low stresses, as seen in Figure 5. The allowable toughness is determined from the actual toughness divided by V2. The temperature exceeds 400 F in the outer region of the reactor vessel during the entire transient, so the toughness is again on the upper shelf, at 200 ksi-sq-rt-in. The calculated stress intensity factor never exceeds 46.1 for an aspect ratio of all = 0.1667, regardless of flaw depth. Therefore, the allowable depth for the fautted condition is alt = 1.0. For the flaw shape all = 0.01, the maximum K = 71.5, so the allowable depth is also equal to the 'hickness. The results for the goveming emergency /f aulted condition:
Flaw Shape (a/l)
Allowable Depth (a/t) 0.01 1.0 0.1667 1.0 o sp =wam neu r, Bruw 2
The allowable flaw depth for this location is then the more limiting result for either the normal / upset or emergency / faulted conditions. In this case the normat/ upset results are governing. The allowables for the flaw evaluation chart are:
Flaw Shape (a/l)
Flaw Depth (a/t) 0.01
.269 0.166,
.355 0.5
.869 Therefore, we see that the allowable flaw depth is very large, regardless of the flaw shape for this location. For conservatism the allowable flaw depth in the chart of Figure 1 has been cut off at alt = 0.2, since the design reference flaw is alt = 0.25, and such a large flaw would be very unlikely.
The only other issue is the potential for fatigue crack growth during service. Since this indication is near the outside surface, and exposed to an air environment, the crack growth during service is negligible, as shown in the table below. Therefore, there is no difference in the allowable depth as a function of service time for this location, and the allowable lines for 10,20 and 30 years of service are the same.
Initial Crack Length After Year Crack Lenath 10 20 30 40 0.500 0.50005 0.50011 0.50016 0.50021 1.000 1.00007 1.00015 1.00022 1.00030 1.500 1.50005 1.50011 1.50016 1.50022 2.000 2.0003 2.0006 2.0009 2,00012 o sop.n<uno, ann.- twe r, thw 2
1'l APPENDIX 1
' TRANSIENT DESCRIPTION FOR THE LARGE STEAM BREAK -
t I
~,
O Goyewantleem11aw Essi for Breedwood 2 -
il I
)
t/x
(
l la 0
4 4
2 S
F 3
2 W
8 0
4 0
8 o
7 6
1 0
O S
1 h
5 5
5 7
4 5
o 1
8 3
6 o
5 t
7 9
1 7
4 3
6 t
e 5
5 5
5 t
5 1
4 t
1 5
5 5
t hgu 2
8 9
4 9
4 7
2 2
8 5
0 5
9 0
6 3
0 5
4 9
o 4
r 8
7 9
t 7
4 3
5 4
8 8
4 1
6 7
3 9
8 4
h 0
5 5
5 5
1 4
1 3
1 5
5 5
5 5
5 T
se cna 8
9 3
5 5
9 2
1 6
7 2
4 t
6 9
6 5
2 3
9 0
0 0
1 3
0 D
5 5
5 3
2 9
3 3
0 3
3 1
s 0
7 9
1 5
4 3
0 0
6 i
5 5
5 4
5 1
2 3
5 5
4 su io 2
3 5
8 3
0 6
4 6
1 2
r 5
4 9
5 4
6 8
4 8
3 0
4 0
a 0
6 o
1 0
8 1
8 4
2 4
4 2
0 5
2 4
6 1
5 t
5 3
2 0
9 5
5 5
5 5
ta ss 8
4 8
2 e
3 1
4 7
5 7
7 3
4 9
8 0
8 r
0 2
2 3
2 2
4 9
4 6
7 7
9 6
9 1
8 2
3 t
5 1
3 1
0 1
8 1
4 1
1 4
2 1
S 5
5 5
4 4
4
/e ru 3
tu 2
2 2
1 3
2 5
5 8
8 8
2 9
4 d
e 0
0 6
7 5
2 3
3 7
8 5
1 2
2 4
5 3
7 7
r 4
1 0
2 1
7 2
1 4
3 2
2 3
2 0
3 2
b p
5 5
4 4
4 4
me T
8 8
3 2
7 8
2 4
5 5
3 5
8 0
9 3
4 3
3 3
6 9
0 7
4 4
5 7
7 7
9 9
1 2
1 7
3 2
3 3
3 0
4 3
8 4
3 6
4 3
5 4
4 4
3 3
70 9
4 1
1 2
9 2
8 6
4 8
2 1
6 8
0 8
7 1
5 3
3 9
7 2
1 9
6 2
0 0
0 3
2 3
4 3
9 5
4 6
5 4
4 6
4 2
8 5
4 4
3 3
3 3
0 4
9 5
2 8
8 3
4 4
3 5
0 0
4 6
3 5
6 3
7 2
8 3
5 0
5 6
8 2
0 1
2 5
4 7
6 5
3 7
5 t
7 8
9 7
6 8
7 6
n 4
3 3
3 2
2 o
itisn p
p se m
o n
p p
p t
p p
p p
p p
i a
k r
o tm e
o aa m
o la r
o se r
o le m
o te r
o r
u o
o m
e u
tn e
s e
o d
e o
tm T
H A
T H
A T
H A
T H
A T
H A
T H
A r
a T
e l
r le B
h t
S n en ei r
il e
sm p
n p
a a r e U
TtS c
c c
c c
c e
e e
a e
e s
s s
e s
s 0
0 0
0 0
0 n
e o
g 0
0 0
0 0
0 r
8 2
3 4
6 6
i a
=
=
=
=
=
=
ta L
c e
e e
e e
e m
n n
n n
o n
i u
u 4
e r
u L
T T
T T
T Y
I I
l 1
i1 1!lll!l' l',l
)
t/4
(
na 2
W o
s 5
1 8
5 T
t 9
t 8
5 1
5 8
4 8
e 2
i 5
4 h
5 1
1 1
1 6
1 t
hgu s
4 4
o 4
2 4
2 4
4 1
r 8
9 e
5 3
4 4
8 3
0 h
0 5
S 5
9 1
1 c
1 4
T se cna 8
s 5
3 4
6 2
t 6
4 3
0 8
0 2
s 0
8 2
8 t
8 7
7 3
4 9
s 3
D 4
e 4
s O
uo e
2 F
5 n
2 e
4 3
T 8
8 5
a 0
4 s
1 7
0 2
8 2
4 5
4 1
0 1
V 4
4 4
ta ss 8
e 3
s 4
3 8
8 O
7 4
8 2
5 5
1 4
S r
0 t
1 2
1 0
2 1
7 2
I S
4 4
3
/e ru t
4 s
3 2
8 F
9 a
2 re 0
5 o
9 3
0 9
3 4
4 7
J 2
s 3
2 4
3 2
p 3
3 3
n re T
a i
t 4
3 5
6 F
d 3
a 8
5 7
8 0
1 0
4 e
3 3
4 3
2 4
3 3
3 3
70 5
9 3
8 3
0 3
8 9
8 8
s 8
7 4
5 4
0 A
0 4
9 4
3 3
3 2
0 r
3 1
9 9
3 0
8 e
4 8
5 9
T 4
0 7
e 5
7 6
4 F
5 4
n 2
2 2
o s
ss g
e l
e p
l g
e o
s p
n p
p o
e se n
p e
e o
se a
k m
le s
r m
e a
m a
i A
T H
A T
H A
c a
o m
A T
H A
T H
A T
s a
u m
sn a
a n
m e
o s
s s
e r
a T
T
(
e s
r e
B h
t S
n e ein re sm n
pp a a r e c
c U
TtS c
a s
a e
e e
t e
0 0
r 0
0 0
o g
o 0
5 r
s 1
1 i
a t
=
=
a L
e e
c
,o m
o e
W o
L v
T T
.l tf G MD I
A - The 10, 20, 30 year acceptable flaw Iseats 8 - Within this zone, the
~
surface flaw is acceptable 20 by ASME Code anc1ytica1 I
C - ASME Code allowable since l
18 i
I 16 l
l l
lli lllh! l(( {
i D - ASME Code allowable prior l
j y
j i
Il i
t j l } l l.
il to 1983 Winter Addendum.
f f
j I
i N
I
'si 12 l
,}
f, I
[
i i
)
.I1
{!
I i
Ei 10 fl I
!l s
- n 4
u!
i L
i gll i
!l uo i t I
diliN !.
f 4
..,,.m i
11i:;IW l h O O:lll
! hh ah MillllWlil!WI4! $
NIMhhhd O
0 0.1 0.2 03 0.4 0.5 FLAW SHAPE (a#)
Figure A-2.5 Evaluation Chart for Nozzle Shell to Intermediate Shell Weld Inside Surface X
Surface Flaw _
tongitudinal Flaw X
Outside Surface Embedded Flaw X Circumferential Flaw w w e m ee e
k c.i 4
Figure 2 Axial Stress Distribution, Braidwood 2 Upper SheH Transition Hydrotest @ 3105 psi e
25 i
t 20 i
+
3 d
I 15 o
w 10 5
t 0
0 0.1 0.2 0.3 0.4 0.5 06 0.7 0.8 0.9 mit, with mA = 0 at outw swface
Figure 3 Stress intensity Factors for Braidwood 2 Upper Shell Transition Weld, Circumferential Outside Surface Flaws, Hydrotest @ 3105 psi 250 4
i lr 200 E
ir l
To
".. 150 I
S
',.. ~
......a1 =.1667 v
S
_.. - a1 =.5 l
~
a/I =.01 mc K-allowable S 100 c
t m
m
<E 250 TO 297 AT t
$.300 4000 SECONDS r
-350 340
-400 0
250 500 750 1000 1250 1500 1750 2000 TIME (SECONDS) l Figure 1.
Cold Leg Temperature *,ersus Time During LSB Transient, T = 557'F (Failed Loop) o i
1.
2-3
in n i
i 50 0
k.
50
_f g
.iOO c
TO 325 AT 150 4200 SECONDS z
200 w
E 226 h.250 g
300 350 327 l
I l
l I
I l
<00 250 m
750 1000 1250 1500 1750 2000 TIME (SECONDS; t
Figure 2.
Reactor Coolant Pressure versus Time During LSB Transient' P = 2235 PSIG o
2-4
1
,u.
1.25 1.00 s<
=
0.75 O
5 o.50 -
t<
E 0.25
~
g d
.05 0
I I
I I
I I
0.25 0
100 200 300 400 500 600 700 TIME (SECONDS)
Figure 3.
Reactor Coolant Flow versus Time During LSB Transient, F = 402400 GPM (Failed Loop) o 2-5
-