ML20214K316
| ML20214K316 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/07/1986 |
| From: | Shepard J, Yoon K BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20214K296 | List: |
| References | |
| 32-1165899, 32-1165899--R, 32-1165899-00, 32-1165899-00-R00, NUDOCS 8612020261 | |
| Download: ML20214K316 (15) | |
Text
F
( ATTACHMENT 4 )
BWNP 20697 (685)
SBabCOCkOWilCOX DOCUMENT
SUMMARY
SHEET a McDermott company l
DOCUMENT IDENTIFIER 32-1165899-00 Tar Fracture Mechanics Analysis of ANO-1 "B" Pumo Case Indication PREPARED BY:
REVIEWED BY:
J. F. Shepard NAur K. K. Yoon NAMF SIGNATURr SIGNATURF TfD, Advisory En ineer DATE TITLr Sup rvisory Enoineer DATE TM STATEMENT:
310 REF. PAGE(S) l!
REVIEWER INDEPENDENCE,
COST CENTEp PURPOSE AND
SUMMARY
OF RESULTS:
To perform a fracture machanics evaluation of the indication found in ANO-1 "B" reactor coolant pump case, according to Appendix A of the ASME Boiler and Pressure Vessel Code,Section XI.
Pertinent Conclusions The flaw indication in the vertical weld in the "B" pump case volute c.rea is found acceptable according to IWB-3612 criteria of Section XI, ASME Boilers and Pressure Vessel Code.
/
THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:
COoE / version / REv CooE / VERSloN / REV 8612O20261 861124 PDR ADOCK 05000313
\\
Ig G
PDR PAGF OF
+
PDS 2103G 3 tt 84)
Babcock & WHcox GENERAL CALCULATIONS a McDermott company 32-1165899-00 Nuclear Power Division o o c. i.o.
RECORD OF REVISIONS Revision Descriotion of Chareces Date 00 Original release 11-07-86 i
l l
l K. K. Yoon 11-07-86 pagpAngo gy atviewto sy OATE PAGE NQ v
,os mum...,
s.a. cock a wiscom GENERAL CALCULATlONS 4 McDermott company 32-115 899-00 Nuclear Power Division o c c. i.o.
TABLE OF CONTENTS 1.
FLAW DESCRIPTION 2.
MATERI AL PROPERTIES 3.
LO/OS/ STRESSES 4.
FRACTURE MECHANICS EVALUATION 4.1 Stress Intensity Factor Calculation 4.2 Fatigue Flaw Growth Analysis 4.3 Safety Factor Assessment 5.
CONCLUSIONS 6.
REFERENCES Appendix A.
Flaw Description PREPAffD SY CATE ys uh/sc 3
PAGE NO.
REVIEWED BY DAtt g
4 PDS.21C33 3 (9 84) s.ncocu s inucox GENERAL CALCULATIONS a McDermott cernpany 32-115 899-00 Nuclear Power Division occ. i o.
FRACTURE MECHANICS ANALYSIS OF ANO-1 "B" PUMP CASE INDICATION 1.
FLAW DESCRIPTION In the "B" pump a radiographic indication which may be slag or incomplete fusion exists at the volute part of the case as shown in Figures 1 and defribed in Appendix A.
The indication is 1.5 inches in length along the vertical weld seam and begins approximately 1.5 inches below the pump horizontal centerline and continues downward.
The depth of this indication is not determined at this time.
Assuming this indication is a deep flaw, this indication is conservatively represented by a through-wall flaw of the same length.
The objective of this analysis is to demonstrate that this con se rv a tiel y assumed flaw is not suf ficiently long to cause any flaw instability and also that there exists adequate margin of safety for one more cycle of operation.
2.
MATERI AL PROPERTIES The ANO-1 pump case was fabricated from ASTM A351-69, type CF8M. Toughnes s and mechanical properties for similar cast stainless steel and weld materials are available frorr. reference 4 and a part of this information is attached in Appendix B.
TFe primary interest is in the weld material which contains this indication.
Type CF8A base metal has slightly higher tensile strengths than type CF8M, however, Charpy values at room temperature are l
approximately equal.
For this evaluation, it is considered reasonable to assume that fracture toughness properties for type CF8M and CF8A weldments are equivalent.
Touchness of Weld Material CF8A Stainless Ste'el Weldment 2 3/8-inch thick compound-butt l
Filler metal - ER308L, E308L 1
l JIC = 1171 in-lb/in 2 KJC =
179 ksik K. K. Yoon 11-07-86 pr.r.u n s.
- s ca 4
..o, o
., _,n,,
n.,,
v 1
~
PDS 21033 3 (9 84)
IBabcock & Wilcox GENERAL CALCULATIONS a Mcoermott company 32-1165899-00 Nuclear Power Division occ. i.o.
Fatigue Flaw Growth Rate Presently a fatigue crack growth rate equation for carbon steel is in Appendix A to Section XI, ASME B&PV code.
The Metal Properties Council proposed a u ni fi ed fatigue crack growth rate equation for austenitic material including CF8M ( refe rence 1) and it is under the ASME committee review for potential inclusion into Appendix A of Section XI.
The proposed equation is
= C F S (AK )
y coefficient, function of temperature (Figure 3) where C
frequency dependency factor, unity for this F
application R-ratio correction factor S
S = 1.0 R<0 S = 1.0 + 1.8 R 0<R<0.79 S = -43.35 + 57.97 R 0.79<R<l.0 At 570 F and R= 0, the above crack growth rate equation becomes as shown
- below,
= 1.8 x10-10 (AK )
I inches / cycle and 4KI in ksib It is noted that all
'where da/dN is in different types of stainless steel can be represented by a single fatigue crack growth equation.
The fatigue data for CF8M in this reference is the same MPC data reported in reference 2.
1 PREPAtto SY o Arg M: 5 u NFC I
01,
,.o
~ o
,,m. o.,
(
i POS 21C3G 3 59 94) sanc cn a wiscox GENERAL CALCULATlONS a McDermott Company 32-115 899-00 Nuclear Power Division o o c. i.o.
i 3.
LOADS / STRESSES The applicable stresses at the flaw location are assumed to be the same as those used for the "A" pump case analysis (reference 5) and repeated here in Table 1.
T6ble 1.
Maximum Stresses at the Flaw Location Normal &
Emergency &
Location Upset (ksi)
Faulted (ksi)
Inside surface 15.9 17.5 Midsurface 12.6 15.0 Outside surface 9.4
-0.6 Since a center crack panel solution is selected for fracture mechanics analysis in the following section, the highest stress valuq will be used as a membrane stress in each category.
l PR, PAR,D
,Y C AT, 075 "Isn 4
P.o. ~ o.
0
POS 21036 3 (0 84 Bat > cock & Wilcox GENERAL CALCULATIONS 3 McDermott company 32-1165899-00 Nuclear Power Division occ. i.o.
4.
FRACTURE MECHANICS EVALUATION 4.1 Stress Intensity Factor Calculation Since the through-wall flaw of 1.5 inch length is rather small compared to the thickness of the case wall (3.1 inches) and the approximate radius of the volute at the weld seam of interest (radius - 9 to 10 inches), a stress intensity factor solution of a center cracked panel is selcted for sinplicity.
Even-though the curvature ef fect is neglected in using the center crack panel solution, given a sufficient margin of safety, it would be adequate to demonstrate the stability of the flaw.
Fcr a center cracked panel shown in Figure 2, approximate solution is given by the following equation (reference 6),
KI = (T k F(a/b)
(1)
F"a/b) = 1 + 0.128(a/b) - 0.288(a/b)2 + 1.5 2i(a/b)3 where For normal and upset conditions, SIG = 15.9 ksi a = 0.75 inch b = 7 inches (essumed)
F = 1.01 KI = 24.7 ksi k...For fatigue flaw growth.
From the fatigue flaw growth analysis for 20 heatup and cooldown cycles in the following section, af = 0.751 inen.
Substituting this final flaw size into the above equation 1, KI = 24.7 ksi in...... Normal and upset.
PREPARED SY DATE ofS n NEL 7
,., s o t
POS 21033 3 (e est Babcock & Wilcox GENERAL CALCULATIONS 4 MCDermott Company 32-115 899-00 Nuclear Power Division ooc. e.o For emergency and faulted conditions, SIG = 17.5 ksi and af = 0.751 inch
- yields, KI = 27.2 ksi k..... Faulted condition.
4.2 Fatigue Flaw Growth Analysis Fatigue crack growth rate is given in section 2 as da/dN = 1.8x10-10(ogy)3.3 where da/dN is in inch / cycle and KI in ksik The fluctuating stress range is 34.9 ksi, however, only 15.9 ksi is tensile.
For the purpose of calculating the stress intensity factor range, twice the tensile stress range will be used to be conservative.
Since the stress intensity factor for normal and upset conditions is 24.7 ksi E from the preceding section, twico this value will be taken as the AKI range.
d KI = 49.4 ksi 6 da/dN= 0.000069 N=
20 heatup and cooldown cycles for one fuel cycle and 240 cycles for design life.
Assuming a constant AKI since the flaw growth rate is so small, the amount of crack growth is da =
0.001 inch ( = 0.017 for 240 cycles).
- Hence, Initial crack dep :h = 0.75 inch PREPAttD BY O Arg tilll7G 7
OAfE P AGE No etvitwt0 SY
PDS 2103 i 3 (9 84) s.ncock a wncox GENERAL CALCULATIONS 8 McDermott Company 32-115 899-00 Nuclear Power Division o o c. i.o.
and 0.75 1 inch ( = 0.767 for 240 cycles).
Final crack depth
=
Since the amount of flaw. growth is very small (0.1% of the initial fl aw depth) for the entire design life, this is acceptable.
4.3 Safety Factor Assessment For normal and upset condition, IWB-3612 requires that KIa/KI > d = 3.16.
Since this material is very ductile (33% elongation - reference 4), it can be assumed that KIa is equivalent to KIc at the operating temperature and the following ratio will be assessed.
KIc/KI = 179/24.7=7.2>k=3.16.
Therefore, this flaw is acceptable.
For the emergency and faulted condtion, IWB-3612 reutres that KIc/KI > b = 1.414 and since KI for the faulted condition is 27.2 ksik 6.6
>1.414.
KIc/MI = 179/27.2
=
Therefore, this flaw is acceptable for the emergency and faulted conditions.
,t,P.R.D BY
- DAY, ys em e
..o
~ e t,....D
P05 21036 319 84)
Babcock & WHcox GENERAL CALCULATIONS a McDermott campany 32-1165899-00 Nuclear Power Division occ. i.o 5.
CONCLUSIONS The results of this fracture mechanics analysis using a very conservative flaw size show that the flaw indication found in the volute weld of the AND-1 "B" pump case is acceptable for the continued operation as per IWB-3612 procedure.
The results are summarized in the following tables.
Fatigue Flaw Growth No.
Initial Flaw Final Flaw Cycles Size (in.)
Size (in.)
Period 20 0. 75 0. 75 1 1 fuel cycle 240 0.75 0.767 design life Stress Intensity Factor for One Fuel Cvele Load Cat.
Max. Stress KI KIC/KI Code Criterion Check A&B 15.9 24.7 7.2 3.16 OK C&D 17.5 27.2 6.6 1.41 OK Stress Intensity Factor for Design Life Load Cat.
Max. Stress KI KIC/KI Code Criterion Check A&B 15.9 25. 0 7.2 3.16 OK C&D 17.5 27.5 6.5 1.41 OK K. K. Yoon 11-07-86 retranto
.y OFS nhtR IO
...~o
. m. o.,
o 1,
(/
PDS-21033 3 (9 84) sancock a wiscox GENERAL CALCULATIONS a MCoermott company 32-115 899-00 Nuclear Power Division o o c. i.o.
6.
REFERENCES 1.
L. A. James and D. P. Jones, " Fatigue Crack Growth Correlations for Austenitic Stainless Steel in Air," ASME PVP Vol. 99, November 1985.
2.
E.
I. Landerman and W.
H.
Bamfo rd, " Fracture Toughness and Fatigue Cha racteristics of Centrifugally Cast Type 316 Stainless Steel Pipe after Simulated Thermal Service Conditions," ASME MPC-8, 1978.
3.
B&W Document 32-1165802-01, "ANO-1 Pump Case Stress," 11/07/86.
4.
M. G. Yassilarus et al., " Stainless Steel Compact Data," NRC Piping Material Program at David Taylor Naval Ship Research and Development Center, presented to ASME Section XI, Task Group on Pipe Flaw Evaluation, San Antonio, June 1984.
S.
B&W Document 32-115797-C0, " Fracture Mechanics Analysis of ANO.-l A Pump Case," 11-07-86.
6.
H.
Tada, The Stress Analysis of Cracks Handbook, Del Research Corp.,1973.
4 PREPAtto BY oAfg E
U N f6 i l
,,v,,w.o.,
o.1,
...~o n
PDS 2:030 3 (9 84>
saack a m==
GENERAL CALCULATIONS a McDermott cormpany 32-115 899-00 Nuclear Power Division occ. i.o.
Figure 1 Flaw Location t0
.\\
s
'c I,,,,,
r 2/
I_
ass R
+4%
760 7
j r 18 J2
. 4
)
i
~m, i.
s tl go
.w g
e s.
O.
- \\
u,{f w
%,. h **' e_m.
\\ O
\\
j O \\
ir y
O
,/ O,,th']
L.
/
o J
=, "
g Q
e
,a 3
, */.,/
/
n ~.
O Q
a'.
v%/
-sis to, str
,/*
"'o a2 gg.
- b0 Ee NOTPRTINENT
[7]
r*/7%b b PitPAGED BY DAff alvit ato SY D Afg P AGI NO l
I PDS 21033 3 <e 34, GENERAL CALCULATIONS mm.cu a wne.x
. uco.rmon company 32-115 899-00 Nuclear Power Division ooc. ' D.
Figure 2.
Center Cracked Panel 0
t t
t 22 9
(
K. K. Yoon 11-07-86 pai,.neo ey II neviewto sv 04te Pact No V
P05-21036 3 f 9 84)
Babcock & Wlicox GENERAL CALCULATlONS 4 MCDermott Company 32-115 899-00 Nuclear Power Division occ. i.o Appendix A.
Flaw Description l
l
,e,*.a.o av 11-07-86 K. K. Yoon l
ys o Iua.
s-
..o,..
L
r 37_- 115 6 89'J _ og
~
ARKANSAS POWER & LIGHT COMPANY Arkansas Nuclear One P. O. Box G08 Russellville, Arkansas 72801 November 5, 1986
//0V06199g ANO 86-14311 Mr. Bill Jones y, # /c.eg Babcock & Wilcox Co.
Nuclear Power Division 3315 Old Forest Rd.
P.O. Box 10935 Lynchburg, Virginia 24505-0935
SUBJECT:
Arkansas Nuclear One - Unit One Revised Request For RCP Flaw Analysis
REFERENCE:
ANO-86-13461 Dated October 20, 1986
[
Dear Mr. Jones:
The purpose of this letter is to revise data transmitted to you by the referenced letter and to request additional analyses of both the ANO-1 "A" and "B" reactor coolant pump casings.
In the "A" pump refined 'neasurements now indicate that the wall thick-ness in the area of interest is approximately 2.6 inches. Other dimensions are unchanged.
In the "B" pump a radiographic indication which may be slag or incomplete fusion exists in the same vertical weld as in "A".
The indication is app ex'mately 1.5 inches in length and begins approximately 1.5 inches below the pump horizontal centerline and contirues downward.
No infor-mation is available regarding depth or throughwall extent for the indication. The pump casing is approximately 3. I inches thick in this area by UT measurement.
We request trat the "A" analysis be repeated using the above revised data.
For the "B" RCP we request that an analysis be performed assuming that the indication extends from 00 to ID and which would show accept-ability for one cycle of operation.
As before, please contact me if you have any questions about this request.
W'f 1.
Q h,. t M
- i; N\\
\\
f4
[
\\
xaiider 6. McGregor,
\\,
/
ABM/jao cc:
E. C. Ewing R. D. Lane R
ayne Toe Scott (BI.W 5 apresenta
)
. ~.,.
s.... -
- -.