ML20214K329
| ML20214K329 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/18/1986 |
| From: | Shepard J, Yoon K BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20214K296 | List: |
| References | |
| 32-1167118, 32-1167118--R, 32-1167118-00, 32-1167118-00-R00, NUDOCS 8612020267 | |
| Download: ML20214K329 (18) | |
Text
. _ _ _ _ _ _ _ _ _ _ _ _ _
( ATTACHMENT 5 )
BWNP 2069/ (485)
@ 88W AWII"'
DOCUMENT GUMMARY SHEET a McDermott company DOCUMEN1 ID[NTIFlFR 32-1167110-00 TITLr ANO-1 Pump Case - CNck Growth Due to Transients PREPAR D BY:
REVIEWED BY:
J. F. Shepard NAur K. K. Yoon NAMF_
h StGNATURF N v SIGNATURr TITLE _ Suoervisory EngjJlear__DATE II/18/8fL-TIILE_ Advisory Fnajneer oATE 11/18/8fL_
^
1 TM STATEMENT:
REVIEWER INDEPENDENC,,$[M Cosi CENTER 30R REF. PAGE(S)
T PURPOSE AND
SUMMARY
OF RESULTS:
a
Purpose:
To demonstrate that only heatup and cooldown transients contribute significantly to growth of flaws found in the ANO-1 pump casings,
==
Conclusions:==
Crack grcwth due to transients other than heatup and cooldown is negligible. See Table 3.
1 1HE FOLLOWING COMPUTER CODES HAVE BEEN USED iN THIS DOCUMLNT:
CODE / VERsloN / REV code / VERSION / RI.V
_ hDY $
8612O20267 861124 PDR ADOCK 05000313 PDR 0
kb PAGL
@T
f o-
)
eDS 21016 3 6 84)
GENERAL CALCULATIONS u awn n e McDermoticomDeny 32.-llb Ttt$-00 Nuclear Power Division ooc. i.e.
REcoR b OF Re vlS 10N,5 Re v' h es c etP rta M
% ATE oo ORIGua 4 L RELEA.SG Ilfitff,6 6
bhb d'II/ b o4tr rastAaro av
'l "l&[N A
'ou exon wo.
m enn e co m
r PD$ 21036 3 (9 34)
Bat > cock & Wimx GENERAL CALCULATlONS a McDeemett company 3 2. - ll 6 H t g-o o Nuclear Power Division Doc. i.o 1.
Intenduction indications were found in During the inspection of the ANO-1 RCS pump casings, the volutes of Pumps A and B.
To demonstrate the acceptability of these flaws, stress and fracture mechanics analysos were performed (see Reference 1, 2 and for both 3).
The analyses showed that large margins exist between Kg and K1 pumps, i.e., unstable crack propagation will not occ0r.
In addTtion, flaw growth calculations showed that the flaw in Pump A would grow an additional 0.039" based on an initial flaw depth of 1.1"; the corresponding values for Pump B are 0.017" and 0.75" respectively.
The flew growth calculations were based on the stressos resulting from the heatup and cooldown transients sinca_ thor.o inadings aro the most severe combination of loads and cycles.
(Notes the flaw growtn calculations were baswd on Z40 heatup/cooldown cycles, which is the number specified for 40 years of operation.
Since some of those cycles have already baen used, this is a conservative assumption).
The question has boon asked whether the, remaining design transients would contribute to flaw growth.
It is the purpose of this document to show that these transients make no significant contribution to flaw growth.
2.
Diccunston nf Deston Transients The design transients and the corresponding number of design cycles aro listed in Table 4-8 of the ANO-1 Safety Analysis Report (Ref. 4).
This table is included as Table 1 in this document.
Dotatis of theso design transients (temperatures, pressures, flows, etc.) are given in the Functional Specif M.ition for ANO-1 (Rof. 5).
For convenience and to provido an overview of the magnitudes of the temperaturo changes involved, the RCS inlet temperatures from the Functional Spocification are shown graphically in Fig. 1.
It is noted that heatup and cooldown (Transtants 1A and 18) are by far the most sevore since the temperature of the reactor coolant varies between 70F and $60F 0
(delta T = 490 F).
All remainin0 transients have dolta T values less than 50 F.
0 In addition, the RCS pressure varies betwoon 0 psig and 2250 psig. during heatup and cooldown. This is by far the largest pressure loading range.
3.
Diennstan of StransesJlua to Transtants In the stress analysis of the pump casing (Ret.6), tho only transient analyzed in dotail is the heatup/cooldown transient (including Rapid Depressurization) since this is the only transient that results in significant stressos.
To determine the af feet of the romaining transient cyclos on crack growth, an estimate will be mado of the stresses resulting from the pressure and tomperaturo variations associated with each transient.
This will be dono as fo11ows:
Cl83 lillR h1 ox,(
eatrAet0 sv
(
REVl[WED SY DAf( _
PAot No
y PD$ 210H J (9 44)
k a=
GENERAL CALCULATIONS 0 MCDesmott Comp,ny 32-IIVfI(9-oo Nuclear Power Division o o c. i.o.
1 The cycles for each transient are dotoririned from Ref. 5.
2.
For each of the transients, the delta T associated with each transient is dotormined.
Only temperature decreasos are considered sinco they cause tonsilo strossos on tho insido surface.
3.
From Ref. 7, Basic Caso 109, the thermal stresses at 3.015 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> into the cooldown transient are:
M Inside:
15.0 ksi l
Midsurface:
11.3 ksi Outsido 7.7 ksi These are oquivalent to membrane (Om) and bonding Ib) stressos of Im = 11.3 ksi
[b=3.7ksi At the time theso stresses occur, the RCS temperature is 250F.
This 0
corresponds to an RCS temperaturo chango of 550-750 = 300 F.
For the various transients, the thormal stress is calculated by mu1Liplylisy Lho vvvidunn oLiosa by the i s.t t o of the tronotont dolto T divided by 300.
For example, for Transient 28, the thermal membrane stress is 43/300 x 11.3 = 1.62 ksi 4.
F rom Re f.
1, Page 43, the pressure stresses at Element 4176 for p =
2250 psi are:
l l
Insido 6.5 ksi Midsurfaco:
7.6 ksi Outsido:
8.7 ksi These correspond to membrane and bending stressos equal to l
l Im
= 7.6 kst l
b
= 1.1 ksi For each of tho transients, the pressure variation is determined from Rof. 5 and the corresponding variation in pressure stress detorniined by rattoing the stressos defined above.
These data and calculations summarized in Tablo 2.
j M5 a i s Is<.
,,,,,,,n o,,
b, PAGE NO.
novitwee eY A t,
S PD$ 210J6 J (9 84) s.w.a a m==
GENERAL CALCULATIONS a Mtoermott Cornpany M~"b iib'OO Nuclear Power Division o o c. i.o.
4.
Finw Growth calculations In this section,,the effect of the transients listed in Table 2 will be evaluated. The procedure will be illustrated using Transient 28.
From Table 2, the variations in thermal and pressure membrane stressos are 1.62 ksi and 0.17 ksi respectively.
It will be assumed that these variations occur simultaneously, which results in a variation of membrane stress equal to 1.79 ksi.
A similar calculation for variations in bonding stress results in a variation of 0.55 ksi.
From Ref. 2, Section 4, the equation for the stress intensity factor for the "A" casing is g+ O M)[TTa/0 Ky = ((T*g M g g where:
Mg = 1.3 Mg = 0.65 a = 1.1 in (initial defect size)
Q = 1.17 y = 1.72 (1.3 C + 0.65 F )
K g
B For Transiont 28, the variation in Kg is Kg = 1.72 (1.3 x 1.79 + 0.65 x 0.55) =4.62ksi%
From Rof. 2, Soction 2, tho equation for crack growth is da/dN = 1.8 x 10-10 (AK )3.3 g
whoros da/dN is in inchos/ cycles dKg isinksif for Transient 2B (1440 cycles), the contribution to flaw growth is a = 1.0 x 10-10 (4.62)3.3 (1440) = 4.0 x 10-5 in.
For purposes of calculating the total flew growth, a threshold of 4Ky - 3 ksik was assumed.
The contributtons of transients with/JKg > 3 ksi % are summarized in Tablo 3.
As shown in Tablo 3, the total flaw growth is 0.063 mils additional to the 39 mils calculatad for heatup and cooldowns in Reforonco 2.
This reprosonts loss than 0.2% of thu growth due to heatups and c ool dow ns and is considered nogligiblo.
GTS t ilf ?;IEL regerito av o4,,
N T
u' etvitwro av orrr race No.
9 PD141056 3 (p 44)
- =a max GENERAL CALCULATIONS
. wo..
u,.,
D IMIIO~OO Nuclear Power Division Doc.io.
Those calculations woro performed specifically for the fl aw in the "A" pump casing, but the conclusion applies to the calculations for the "8" pump casing as well.
5.
Effect of Hvdrotant on Flaw Growth According to Ref. 5, Para. 6.13, a total of 15 field hydrotosts are specified for the reactor coolant system excluding tho soconda ry side of the OTSG (the 35 cycles specified in the SAR are the total number of OTSG secondary side hydrotests).
'Section 15 of Ref. 4 states that the hydrotest pressure is not to exceed 2750 psi and that the tosts are to be conducted at tomperatures greator than 215F.
The stresses at Element 4176 for p = 2250 psi are shown on pago 4 of this document. The stresses for p = 2750 psi are:
Inside:
7.9 ksi Midsurface:
9.3 ksi Outside 10.6 ksi The membrane and bending strossos aro
= 9.3 ksi
1.4 ksi These stresses result in a stress intensity factor of Ky = 22 ksi h for a
1.139 in. (final flaw sizo).
From Appendix A of Section IX of the ASME Boilor and Pressure Vessel Code, tho value of K at 21SF is greator than 120 ks1Rassuming RTNDT = 72F).
Use of a
carbon stoe pmporties is conservative for stainless steel l
l Tho safoty margin for hydrotest is then l
l Kya/Ky > 120/22 = 5.4 > 3.16 For AKg = 22 ksi % and N = 15 cycles, the crack growth contribution is a = 1.8 x 10-10 (22)3.3 (15) = 0.0001 in 6.
Co nclu s ion s Only heatup and cool dow n transients (Transients lA and IB) contribute significantly to flaw growth.
All other spocified design transionts have a
"*ul i u t t,1. or roa v..
u.o r i.. 1 riu...
H lS PREPAttD BY pagg i
II I noviewtD er oatt e40e No.
7.
PD121036 ) (9 34) em cu a muax GENERAL CALCULATIONS
. ueo.v. mmens 32-UWi 8-00 Nuclear Power Division o o c. i.o.
Rafararras 1
B&W Document 32-1165802-01, AND-1 Pump Case Stresses (NSS-8).
2.
B&W Document 32-1165797-00, Fracture Analysis of ANO-1 A Pump Case Indication (NSS-8).
3.
B&W Document 32-1165899-00, Fracture Mechanics Analysis of ANO-1 B Pump Case Indication (NSS-8).
4.
ANO-1 Updated Safety Analysis Roport, Facility Operating License Nos. DPR-51 and NPF-6. Docket Nos. 50-313 and 50-368.
5.
Specification CS(F)-3-92/NSS-8/0372, dated 3/22/72.
6.
B&W Document 33-0210-03, Pump Case Analysis for Consumers Power Co., Midland Units I and II, Byron-Jackson Report TCF-1023-STR, Volume 2, Rev. C.
7.
B&W Document 33-0086-00.
This document contains the microfiche for Ref. 6 above.
~~
"E i
r4o: wo.
etv eweo er 04:e
7AdlE /
32-li bli18 - 0 0 ARKANSAS NUCLEAR ONE Unit 1
(
Table 4-8 TitMf 3IENT CYCLES Transient Design Number Transient Description (ASME Category)
Cycles 1A Heatup from 70 F to 8% Full Power (Normal) 240 Cooldown from 8% Full Power (Normal) 240 2
Power Change 0 to 15% and 15 to 0%
(Nomal) 1,440 3
Power Loading 8% to 100% power (Normal) 48,000 4
Power Unloading 100% to 8% power (Normal) 48,000 5
10% Step Load Increase.
(Normal) 8,000 6
10% Step Load Decrease (Normal) 8,000 7
Step Load Reduction (100% to 8%
Power)
(Upset)
Resulting from turbine trip 160 Resulting from electrical load rejection 150 Total 310 8
Reactor Trip (Upset) g Type A 40 Type B 160 Type C 88 Trips included in transient numbers 11, 15, 16 17, & 21 112 Total 400
)
9 Rapid Depressurization (Upset) 80 l
l 10 Change of Flow (Upset) 20 11 Rod Withdrawal Accident (Upset) 40 l
12 Hydrotests (Test) 35 13 Steady-State Power Variations (Normal)
=
l 14 Control Rod Orop (Upset) 40 15 Loss of Station Power (Upset) 40 16 Steam Line Failure (Faulted) 1 17A Loss of feedwater to One Steam Generator (Upset) 20 k
178 Stuck Open Turbine Bypass Valve (Errergency) 10 gps u (tg (F(,,
g q/gf gp _
(
4.6-13
M ot& I (CodriMOED) 3 2-l { b } l l 13 -(> O c/
ARKANSAS NUCLEAR ONE Unit 1 Table 4-8 (continued)
Transient Design Number Tynnsient Descript.fon (ASHE Catcoorvi rye?~.
18 Loss of Feedwater Heater (Upset) 40 19 Feed and Bleed Operations (Normal) 4,000 20 Miscellaneous A (Normal) 30,000 Miscellaneous B 20,00g Miscellaneous C 4x10 -
21 Loss of Coolant (Faulted) 1 22 Test Transients (Test)
HighPressureInjectionSystem 480 v
Core Flooding Check Valve 240
[
23 Steam Generator Filling, Draining (Normal) g Flushing, and Cleaning k
Steam Generator Secondary Side Filling 0
Condition 1 120 1
Condition 2 120
(
qe y '2 Steam Generato,r Primary filling Condition 1 120
}v Condition 2 120 hI Flushing 40
- o d A-Chemical Cleaning 20 W
540
~
24 Hot functional Testing (Test) 1
+gs 4.6-14 M
U/(r/Pd
u h
I I
jg
,t.,.
..-4_.
k b
N5 O
W h o' 7
d0[ ! ?
O. oi)
) !
$2 3
ef
- l4M, d
6
+
4
+
O i
i (O
A_
__7..b
{k A
g4 h
7 T
T
~
7 h" T'$
+2, Y
".r "o* "J dy.
g e 3
2e n-of o f q
l u
+
i
+
i i
J t
i re IM l
_w
~
s e
4 g
$j 0, y oU g
sh C
oh o
o o
o O
g o
n(
u-os g
e u <
- q sq u
u v
+
o
+ '
4 i
+
+
+
e 4
+
l_.]
Es i
C I
J
{.
g m
A A
+
+
W a
a
'g F
i y
8
~i e
a y
a rg e
g
$l I
o' o.
o o'
o' o
e q
oj a
3 j
3
>>'E i
i v,
!.f J...
l i
d
=!
t in!
a c
s w ' 'I l
s 2
t; l '1 ' i f l v
i
+
J to d
e o
l,:
)p J..
j ll g_
I ll "3
lI l ;;
i Id 2
tl I
. i i4,s I
a d'f
]k~'
g; jl~
i i
l l[ '
i 4
f..
$m n,
'u
.-l a
a
'I 4
i w
d 4L l ',
N n
7 W
W i r l1 tw 1
l I,I w
I i
.a i
e i
h I
3 0
y a
c o
o l
i c
1 7
9
?
o 4
o o
y 3
s N
V m
7
> A, T
,o 1
M
,3 l l q
~
l,i l
l g
.... ____ p l
l k.
e,
.r b,
o A
N 4
d
~
l Po3e to
. _R_
i b'
l
'gf5 ulte(g IJV IlI ?
x y__
ffh
F
+
i i
i l
l i
td D
O 0,
ol ad a
a' k - ed n a g
i o
6,2 i
i o
t ' ll ;> +
e -
4 d
O t,
4 g
g g
M l
m9 ln h
o O,
n
+
y e
v n
~
' 'I !
~
I C
o ya o
a t, do.
f4 0, o, o
{
u w
I i u_m ed a,o' do a
o-e I
y 4
lil g
~.,
mr f s' i
l
ii H
4 a
+
t
'\\
l ll i!
I, l
.. g.
l y
e<
r3 41 e
o' o
0,
u n ;
- 'r J
t Ir y
i N,$
d d,
'0 c'
o' l
A
.n r,
I
'I
! t I
I i
+
,j I
(
r 6
- i.. !
. j...
i: ll:.
u p
C i
+
.a.
i I
r g
.o (5
I II !!
lll r
o, g.
e-
! ; i:
< )
l N ! 5 ' ' 0) dgn 4 "-
4 3
j L.
j-,j!
t.I l,
I l
l l
l i
J i
4 i
I I (
l i
lI lij 4
?
l i
l s
a '
in hl I
i r
i l
Qi e
i e
g 6
l
)
B u
o o
o 0
o I
l 1 1
H s
2 7
d 1
0
,i i
l i
i i
i i
9
.1
.I.
w I
M.
- 1 to e
P9 0 g
.p.
g QFb H i l3 lC\\b fQ-UI(lYlf g
\\b C95 21036 J (9 44 sabcock & Wilcox GENERAL CALCULATIONS
. w o.,-~.-,
31 - " G " 9 - D Nuclear Power Division o o c. i.o.
Tablol Tranntant Aly(ks1Rn) b (mtis) 28 4.6 0.04 3
2.6 4
2.6 5
0.6 6
0.6 7
3.1 0.002 8A 2.5 8B 4.7 0.005 BC 5.8 0.005 9
a 10 1.5 11 9.0 0.010 13 0.32 14 1.7 15 2.9 17A 4.0 0.0003 17 8 5.3 0.0004 0.063
- Transtont 9 is included in the heatup/cooldown transtent analyzed in detail in Ref. 2.
b 0
I b
PareAtto ev o4, D %, d[!r/r6 it
.r m wro., -
q Babcock &Wilcox en.mn.
n.n>
Nuclear Power Generation Division GENERAL CALCULATIONS t
Rcs PUMP TRAust ERTs TRA%
Fic, ORE g cyct.Es gy.]
FLoO /TE mP vs T1mE gep/dee=o/sd gg gg_g (Z4ol 45 "!'#'
I 8 O" 4.9B%
g gp
(* F)
REATOP 6w'M b-d d,
6 o
Time. (Mh I
o /zsci
.J
\\%
\\6-\\
Rcs
,4 (A'} o)
SEhD ti g6 e
COOLboldra
\\
000%')
l
[k"S o
2 4
6 6
to 12 575*
Es.coc Gen
[
7_R Z A-1 m
b440) 2o gio.
l Fi c,,
i 1
o 1
2 3
meano n Ws itji&I?(,
m.52 - U ca-t i e.-oo M
m,.
Ill(@I. o.
13
Ii Babcock &Wileox
...m n.> c..n
..n.....,
e' nuci..r row.r a.n.r.uon oivision GENERAL CALCULATIONS TRe u 1: wage s,s'/
P mo-cams Nest-ee.om um
- C5 7 e, 2.6-1
[1490 )
o i
___ z s
(ad s,s Rcs BB,@ STA p
551 /
555 (48,000) t's o
10 20 so (mM 5%*
4 4-1 553' sss' (12,04 o
10 to so (md 559L 55(,
.5 5-\\
se ks
$$<ooo em
( 6,000) fus.t(con'r) o I
7.
(mso) sk lu,
_. n-um e-op m.m..,
ws
,a
(#7...tiltM Li 9
_1
IS Babcock &Wilcox
- a >" > o u' 3
wucisar Power Generation Division GENERAL CALCULATIONS TPMs make.
sus' A t'tct Es 334
?
88.0 N
55"),
('o
(.o- \\
(rooo)
O l
4 3
(roid 5W RCs 98.000 G hn 7
I-i sse sww sse (sto)
~
s o
5
\\0
\\S (mul 5se 2 Aw,sss-ac.s 8R 8-1 sem % m a-tso ea "To To l
d40) l
%d R 6
o l
sw sse Rt.s espe arm 8o s-4 sse s, -
(\\bo) ris 1 (co dr) s O
i1 2.
3 N-mw u... -
nFS
__. n _.nl t/EL
=,,, 32-11 s a u s-o o o
W.
olt,/rf i n - -
lg Babcock &Wilcox
- i>>ar> o.ai
(
wuci..r row r o.n.r. tion oivi. ion GENERAL CALCULATIONS I
5' '
TRaus F\\wRE g3 as,coe m A C.YCLES 30s-
- 554, BC B-7 (BS) e o
i 2.
3(md 5 6 D' -
RCS 88, ooo 69t<\\
9 S-I wo' A t.
o 5
to ts (mvh U
sss' sy~
\\O
\\o-1
\\sigss ooo 6m (7,O) 5% ?.*
RG 88,000 69m 57 5 -.----
1l 1l-I (A, o)
Fin I fcon'r) l q
v o
1 2.
S 4-(,ma) restano o 0FS
..n u a IFt.,
32-tit vu e-e m,,..
en -in, _ ____"_ _.
(D h n/ F[96 l,,,,,.. j_b__
i7
'a >>an > n n>
Babcock &Wilcox GENERAL CALCULATIONS Nuclear Power Generation Division M'
TRRMS FMURE sss*
3CfC.LES ss0 qcg gg,gg, cr,m 14 14-1 (40) o i
2_
3 4
(mud g
sW pcs
\\s
\\s-\\
es.o.a q p asa ng dea a m
(
aJ A u
e s s s'.
"'Is s
gq g_ g
&#)
0 O
I 7
3 (eQ b _
RC6 66,000 @
- 5(/$
5f,0 T7E>
%\\
54*/
sw
\\0)
Fic..t (c.odr.')
l O
2 4
(,
(md)
=,.. n vs w, _
vitJgp c
n-is a ti e-w b m ' S.
esos o. c }
l
N-
ig :
Babcock &Wilcox
'a>">n'l nuci.ar Power Generation Division GENERAL CALCULATIONS,
TRROS FIGOAE sso RC S 90.000 GPM
( CYCLES FG K6 $ BLEEb bo65 Mor A w F acT" Rcs T1nPsAArage 0
1 O
550 RCS 86 N bD
?D no scre.cT
.a PoHP
( SEE d 6TE SELoW) i
%./
O TR eu S W.4T 20 To Acco uoT roR. udguned TRAQstpoTs o4 TME M Ar.E O P, S URC-E, AOb SPRAY t t o E.S Aub McEEWS,TM6 Facto o.cc, KiscaccAusou.s Ta AcsssuTs Aat. s p eciF iEb ',
A. M An il?
Ftous taoes vii.og u n an gt. es.ow ro g c,e n Fo R.
Is uso urt s,iu e.as AsEs To 3R GPM 'Fon C H io oTt
)
A4% T'REA
'b g c ky,A s c s t u ObaHAL F Lond. (3 o,co o c 4c d No EFNct' oc T%E Pon?.
G. SeRav s s Asc uHe.w to ce AcTuATeb
^^ S kaw" '"
Fib. 20 2.
oF RE F, S~. M s.
H As, uo F FE CT o d THE Po HP.
(.ao,ooo c 9 c.E E )
c., K AKE o?
Flow b ti.o PS F%.o m n o e. n A t.
F t.o u) W G. G P F FoR. o.5 MiA.3 IMcREAsEs m,lo GPH gor o.5 Hio.3 Aud n.E T u e.4 S To uoRMAL Ft.o e. ( 9 x, t o'* c N ctcr. )
hio E FFEC.T ou TH E P ts HP, g,7 g eq
,,,m,, _ __
!E 8*c
- 32 u t.w 8 - oo
_ om u
ggieb ir
...