ML20039F413
| ML20039F413 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 01/10/1978 |
| From: | Taylor J BABCOCK & WILCOX CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201120434 | |
| Download: ML20039F413 (25) | |
Text
_ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ -
Babcock &Wilcox eo ercenereuencroup t
l P.O. Box 1260 Lynchburg, Va. 24505 Telephone:(804) 384-5111 p'. 'e.
January 10, 1978
$.t Ws
'N
[D /.
.,s
!$ N il (f g 2 3?978 h$
Mr. S. A. Varga, Chief Light Water Reactors Branch No. 4
~,, M*f @ tusa Division of Proj ect Management Office of Nuclear Reactor Regulation
~,.
7 U.S. Nuclear Regulatory Commission b%'b/
g6 '
Washington, D.C.
20555
SUBJECT:
ECCS Analysis of B4W's 205FA NSS at 3600 MWt and 103,500 gpla RCS Flow
Dear Mr. Varga:
The attached ECCS analysis has been performed to verify the conservatism in the operation of the B4W 205FA NSS at the above conditions.
This analysis was performed with the August, 1977 BGW ECCS Evaluation Model as contained in BAW-10104, Rev.
3, "B4W's ECCS Evaluation Model."
This analysis has been performed utilizing a containment backpressure (Figure 12) somewhat lower than that previously used in BAW-10102, "ECCS Evaluation of BGW's 205FA NSS."
This lower backpressure has been used to ensure that the analysis envelopes all B6W 205FA applications which reference the above topical reports.
The Staff and B6W have discussed the analysis as desc ibed I
in the attachment to this letter during a meeting on January 9, 1978, concerning WNP-1/4.
BAW-10102 will be revised to incorporate the attached analysis.
If there are any questions concerning this submittal, please contact me or Henry Bailey (Ext. 2678) of my staff.
Ve truly yours, CfhMN
/Ja'mes H. Tayloy' Manager, Licensing JHT:dsf Attachment cc:
R. B. Borsum (B6W) 8201120434 e20403 780240203 PDR FOIA MADDEN 80-515 pDR The Babcock & Wacca Company / Estabt;shed 1867
+
Introduction 2
The 8.55 ft double ended pump discharge (DEPD) break, C = 1.0, is the largest D
possible break at the pu=p discharge.
It was chosen for s reanalysis which envelopes the Washington Nuclear Project because it is the w:rst break identified by the spectrum study of BAW-10102, ECCS Evaluation,of B&W's 205 FA NSS.
The 6-foot elevation was chosen as the location of peak power because it has the highest allowable linear heat rate.
Summary and Conclusions The peak cladding temperature for this analysis as shown in Figure 1 is below the 2200F limit and 62F below the peak cladding temperature or BAW-10102.
The maximum local oxidation is 4.27% which is below the value reported in BAW-10102. Since individual cladding temperatures and oxide thickness are less than those in BAW-10102, the analysis therein confirms that the whole-core metal water reaction will be less than 1%.
The max 1=um blockage of 59.1% indicates that the core remains in a coolable geometry, and as explained in BAR-10102, a clear path to long term cooling is eatablished.
Method of Analysis B&W's system of computer codes and related analytical techniques as described in BAW-10102 is used for this analysis with the following exceptions. The computer models were modified to envelope the Washington Nuclear Project parameters and also to be consistent with the most recent evaluation model. They were revised to be consistent with a power level of 3600 Mut and a system flow rate of 103,500 gpm/ pump. The effects of the following model changes which were not considered in BAW-10102 are included in this analysis:
1.
A 0.25 psi penalty in the cold les pressure drop due to HPI injection during reflooding as required by letter from J.F. Stolz to K.E. Suhrke, January 16, 1976.
2.
THETA revised post-CHF heat transfer logic as approved in letter from S.A. Varga to K.E. Suhrke of February 18, 1977.
~
3.
CRAFT 2 revisions approved in letter from S.A. Varga to J.H. Taylor of May 13, 1977.
CRAFT In order to model a core power level of 3600 MWt and a system flow rate of 103,500 spe/ pump, the input parameters pressure and enthalpy are revised for each control volume. Other input items which are different include the following:
path flow rates, ref erence heat flux, peaking factors, average fuel temperatures, steam generator heat loads, and initial pumn parameters.
The hot channel is operating at a peak linear heat rate of 16.8 kw/f t as in BAW-10102 with average of 5.654 kv/ft.
l
___m__
REFLOOD In this analysis, the k-factors for paths 7 to 9, 8 to 4, and 9P to 4 are adjusted to account for the increased pressure drop of 0.25 psi due to HPI flow as mentioned above.
These adjusted k-factors are shown in Table 1 along with those of BAW-10102.
CONTEMPT The CONTEMPT code is used to verify that containment pressure history utilized in the ECCS evaluation is equal to or less than the minimum contain=ent pressure history which would result in the Washington Nuclear Project building. The curve utilized in this evaluation was chosen to conservatively bound (be lower than) all present B&W 205 FA plants. Figure 12 shows the family of curves of containment pressure for the B&W 205 FA plants and the resultant bounding curve. Each pressure history was developed using system parameters and initial conditions which will conservatively underpredict the pressure history. The procedure is defined in BAW-10104, Rev. 3 "B&W's ECCS Evaluation Model," and most of the particular pressure histories have been previously reported in BAW-10102, Rev. 2, "ECCS Evaluation of B&W's 205 FA NSS."
Containment pressure is only used during refill and reflood, which is why most of the pressures are only shown from 20 seconds on.
Figure 13 isolates the comparison of the Washington Nuclet* Project pressure history to the history utilized for this ECCS evaluation.
As can be seen, the evaluation history is well below the plant specific calculation. The plant specific calculation is reported and justified in Appendix A of BAW-10102, Rev. 2.
THETA A new version of THETAl-B is used which contains revised post-CHF heat transfer logic which does not allow return to nucleate boiling as mentioned above. The hot dimensions of the fuel and cladding are revised for each axial level.
Significant parameters are shown in Table 2.
Results Figure 1 shows the cladding te=perature transient for the ruptured and unruptured levels. For the unruptured level, the initial temperature is approximately 710F.
Af ter the initiation of this accident, this temperature rises and falls as core flow either decreases or increases in magnitude with change of direction. Af ter initially oscillating, core flow becomes positive at 1.2 seconds, increasing in cagnitude until 3.6 seconds. After that, the flow decreases again until it goes negative at 10.7 seconds and remains negative until the end of blowdown at 22.2 seconds. The temperature rises until 12 seconds, then f alls until just before the end of blowdown.
As blowdown ends and adiabatic heatup is occurring, the temperature again rises. Rupture occurs at 26.2 seconds. Adiabatic heatup ends at 34 seconds. The cladding temperature in the ruptured node peaks at 1818F at 40.5 seconds (Figure 1).
For the hottest pin, the transient pin pressure l
is illustrated in Figure 11, gap h eat transfer coef ficients in Figure 5, and surface heat transfer coefficient in Figure 4.
_________._________,___8.._____
t f
The temperature in the unruptured level continues to riss to a maximum of 2064F at 98 seconds and decreases thereaf ter as the heat stored in the fuel (Figure 6) decreases as the increasing heat transfer coefficients from ECCS fluid cool the pin.
Additional data are given in Table 3.
Referer.cas 1.
R.J. Lowe, G.E. Anderson, Jr., and B.M. Dunn, ECCS Evaluation of.B&W's 205-FA NSS, BAW-10102, Rev. 2, Babcock & Wilcox Company, December 1975.
2.
Letter from J.F. Stolz to K.E. Suhrke of January 16, 1976.
l 3.
Letter from S.A. Varga to K.E. Suhrke of Feburary 18, 1977.
4.
Letter from S.A. Varga to J.H. Taylor of May 13, 1977.
~
List of Tables 1.
REFLOOD Code Re istances 2.
Comparison of Significant Thermal Response Parameters 3.
Summary of Break Data O
O e
SG 9
e e
O e
S O
-= =.-
O m
1 l
1 e
9 9
~
Table 1 REFLOOD Code Resistances m
g g.
BhW-10102 WNP Path Area, ft2 k
From To, k
7 9
5.858 13.74 14.52 7
8 4
4.276 0.82 -
1.973
~
9P 4
8.552 0.82 1.807 c
N e
6 n.,.
s.
t Table 2 Comparison of Significant Thermal Response Parameters BAW-10102 WP Core Power, We 3800 3600
' Core Power to Fluid, W e 3771.348 3572.856 Radial Power Factor
- 1,67 1.76 -
~~
Axial Power Factor 1.7 1.7 at 6-foot level 9
O e
e
\\
n
- \\
I 1
]
i q
e 9 s...s e
e a
b 6
4 Table 3 Su=sary of Break Data Break Location PD Break C D
Peak T unruptured, F 2064 Time, s 98 1818 Peak T, ruptured, F 1
Time, s 40.5 Rupture time 26.2 CFT Actuation Time, s 15.39-End of blowdown, s 22.2 End of adiabatic heatup, s 34 Lower plenum and downco=er water level
_.s at E0B, f t
.177 Mass of water in reactor at E03, lbm 1019 Full power see at E0B, a 1.782 Local H-W reaction, %
4.27%
Hot channel blockage,' %
~~~
59.1 Cont, pressure at time of peak T, psig 18.5 e
e 9
l l
List of Figures 1.
Hot Spot and Ruptured Node Cladding T.mperature 2.
Core Flow Rate 3.
Core Reflooding Rate 4.
Hot Spot Surface Heat Transi er Coefficient 5.
Cap Heat Transfer Coefficients for Hot Spot and Ruptured Node 6.
Hot Spot and Ruptured Node Fuel Average Temperature 7.
Hot Spot Coolant Temperature 8.
Hot Spot Quality 9.
Reactor Vessel Pressure
- 10. Core Power
- 11. Hot Pin Internal Pressure
- 12. Containment Pressure Family and Envelope History 13.
Contain=ent Pressure Assumed in the Analysis With WNP Specific History
- 14. Downcomer and Core Water Levels 9
e L.
e
Figure Computer Code Number Version Name Version Date Run Nase Run Date N'
1 THETA 1-B, Version 6F-1 6/14/77 WT223V5 1/6/78 2
CRAFT 2, Version 8.1 5/03/77 PA223UJ 6/20/77 3
REFLOOD w/ Loops, Ver. 1 12/20/74 WP223Ny 1/6/78 4
THETA 1-B, Version 6F-1 6/14/77 Wr223V5 1/6/78 5
THETA 1-B, Version 6F-1 6/14/77 WT223"5 1/6/78 6
THETA 1-B, Version 6F-1 6/14/77 WT223VS 1/6/78 7
THETAl-B, Version 6F-1 6/14/77 WT223V5 1/6/78 8
THETA 1-B, Version 6F-1 6/14/77 WT223V5 1/6/78 9
CRAFT 2, Version.8.1 5/03/77 FA223UJ.
6/20/77 10 CRAFT 2, Version 8.1 5/03/77 PA223UJ 6/20/77 11 CRAFT 2, Version 8,k 5/03/77 PA223UJ 6/20/77 12 CONTEMPT, Version 21 9/01/76 PC223BK 7/14/77 CONIEMPT, Version 15 11/15/74 VP2239Z 4/07/75 CONTEMPT, Version 15 11/15/74 WPPSSGM 7/09/75 CONTEMPT, Version 15 11/15/74 PGE0lMF 7/28/75 13 CONTEMPT, Version 15 11/15/74 VP2239Z 4/07/75 CONTFMPT, Version 15 11/15/74 WPPSSGM 7/28/75 14 REFLOOD w/ Loops, Ver. 1 12/20/74 WP223NV 1/6/78 t
9 i
i i
l 1
g 9
0 2100 2064F t
1900
~
1818F
.:p:
p
/
\\.
1700 s. %. --
N.
w N.
5 1500
\\.
i
' l 1
N.
1
=.
i l
N l
t 1300
~
=
G Hot Spot 9
Ruptured Node 1100 900 e
700 t
I t
I 0
25 50 75 100 125 150~
Time, s HOT SPOT AND RUTPURED NODE CLADDING TEMPERATURE m m_ _ _
m
(
i W
- w..
15 1
/
10 -
n
=
0 u.
5 Cc co w
O L
f
=
i C
0 4
=
5 I
u.
1 5'
k0 I
I L-1 0
5 10 15 20 25 30 Time, s CORE FLOW RATE Figure 2
12 iH.
8
~
~:
s o-4 i
i M.
t$2 g ~,
g 2
m
-4
- 8 I
I I
I i
D.
25 50 75 100 125 150 175 Time After E08, s 1
i i
i CORE REFLOODING RATE Figure 3
.?
O e
p e
e a,
e e
9 e*
e e
e b
9 a=
=
W.
M 4
eau e
6 dum 8
m
=
=
M 4
ed sud e
sul 6
m M
W e
pe4 WE e*
N 2 I g!
4 R,
as eie I O t
J
=
J O
g-
="
l$
4 m
3 o
=
me S
e en udD g
N en M
too
~
=
-=
e se e
M s4S I t'I O
clus M
e 4
e 4
10'I
"=
=
4 m
W GUI 4
eG 14*3 g
13 36
$4 12 to 100 126 144 182 ISO f te s e
NOT SPOT SURFAC[ H[4I IRANSFtt COEFFICitNT Figure 4
1000
^
oT J
n 800
~
R
=
G 5
600 5
/
0 I
5 400 l
Hot Spot j
l
_ _ _ Ruptured Node O
I Si i
e i
200 I
p m
~~
3 i
i I
Ip,._________________________________
0 t
i i
i i
i I
0 25 50 75 100 125 150 175 200 Time, s GAP HEAT TRANSFER COEFFICIENTS FOR HOT SPOT AND RUPTURED-NODE Figurs 5
,,_.__.__m
/
o o
/
w
=
m o
e z
/
2 o A
=
a w a
/
w m w A
a w W
a.
m o o
= <
K
=
a w I
z >
a
~
n e-.
a a w A o M A W
E N
=
z e=*
a a
o a -
3a a
o z =
=
1 s
t o
i o -
1 ll 1
l
\\
\\
o i
\\
o
\\
\\
o
- v 1;.
1 4
o t
- m
/
1 s
1
\\
/
/
i 1
o l
g 8
8 E
e og i
o a
m o
l g
n
~
m l
Jo'ajnlejadcal s3cjaAV l203 l
i 18 e
16 e
8 h
- 5
'k o
m' 12 O.
S 10 B
2 2.
G$
8 l
2 O
O L
6 l
4 l
2 e
i i
i L
0 25 50 75 100 125
,150 175 Time, s HOTSPOTCOOLANTTEMPERkTURE l
d Figure 7
1.0 1
0.8
,q ;.
-Q 4
0.6 2
E es 0.4 W
0.2 0.0 I
I I
I O
25 50 75 100 125 150 Time, s l
HOT SPOT QUALITY Figure 8 a
,_m
35 30 m
.'/. l4. : Wi 25' n
'a N
i l
v 20
.=
E 5
=
an in" 15 n
10 5
O i
i i
i i
O 5
10 15 20 25 30 l
Time, s l
i REACTOR VESSEL PRESSURE Figure 9
I 1.200 1.050 -
O
.900 -
~
O jfi:
' L) a as n.750 -
a.
s.
O
=ce as SB E'.500 -.
a wa
.450 --
l 1
.300 --
.15 0..
0 0
5 10 15 20 25 30 Time, s CORE POWER
,j 90 n
~a
~
88 M
v a
i yt.il:
- .0 d
86
=
M M=
w hl a.
-a C
U 84
.5
_C.
a.
a W
82 80 78 I
[
l 76 0
5 10 15 20 25 30 Time, s HOT PIN INTERNAL PRESSURE Figure 11
o=*
I a
6 a
e oc M
c1 u.
A = o L.
t w.-
t.'.
- /
=>
m m s
/g m =
I m-s
./,!
o w w s
d
=
a.
a.
o
//
o s
a l
.,.f e-w z>
8 w z
..I
= w
/ lI s
./
./
.n:
- f e-
./
g x
o
- /
/. /
u
. /.
..'s V. ) '
g..f I
\\
=.
=
o C
~
a.
oaw w
v m
m
>=
3
-c z<
m oW m
o e-z s
n.
E 5
(
o m m o
o a w o
a o
o a.
a u
o n.
=
u.
w z a w o
n.
. a m
's i
l l
1 s
l i
l 1
I s
i l
I
' s I
i o
o o
o o
o e
o m
~
c.
I 3!sd 'ajnssaJd luacuicluo3
~
I H Y T R O
N T 4
S I
I D
D H E
E S
M C C
U U
I S
S F I
F NI )
S I
OSE A C I
E C
YP I
E DTLO E P R S P
OPAL S
OMNE U
LUAV S P P
FS N
S N N
ESNE E W R
W RAI (
P H T
4 T
I f%
N W E
g M S N
I 0
S I
A Y N
0 T L 0
s%
N A 1
O N N
C A
~
\\
\\
s
\\
\\
em i
T 0
0 1
0 l
5 0
5 0
5 0
2 2
1 1
$a. EM5.n3o I'
i
I.
<!II
~
n== 8hOo a
8 0
8 6
4 2
0 1
0 0
2 S
n L
L J
E V
n E
n L
nvr 5
R u
i 7
E c
1 T
AW E
R O
0 C
5 i
1 D
N A
_~
R E
M O
C N
5 W
1 2
O 1
D s
r 0
0 e
i 1
e m
0 m
i C
R T
N 0
0
[
e r
5 0
i 7
C 0
5
. i-s e
5 i
r 2
,j
+
J 0
0 0
0 0
6 2
B 4
1 1
M-
._,,7,a S 2 aeooCEO3 s
C
- I
- '
o t
l-l\\,
l' I
l' l,
v.s. Nue:.z.An nEcut4 Tor:V ccuusssic ocem ltiliticaw125 W9/CA 4.
(Mel NRC DISTRIBUTICN pon PART 50 DOCKET MATERIAL FROM:
cATR or occuusMt 70:
Babcock & Wilcox 1/10/78 ',
Mr. S. A. Varga Lynchburg, Va*
oars maca vec James H. Taylor 1/16/78 cCarrsa CuoTontzso PROP (NPUT Po RM NUMS4n CF CcPitS MECS VEo 2 tis:toman.
EtInct.assarino i sie,.ww Ccorv auctosuna
' cucniPTica i
I ECCS Analysis of B&W's 205%SS at 3600 Wt & 103,500 gpm RCS Flo*....
t l
(1-P)
(23-P) l FLANT MAME : RJL 1/24/78 817~ At t). f.AAl t'A* /kY70
/ 4 we d.
l FOR ACTICN/INFOl;MATICN ENVIRONMENTAL es..-c FAPJ644t O ASSIGNED AD:
'7.
WOORE fLTR)
A ce%. u ine_
/
noANCM CHIEFt Mf7f%/9 BRANCH CHI.EF:
- d. f.A.A a Jo s own Tec* VriACTR' I
/'Po0JEC* MANAGER:
r Tr*.
iss**
t.IC. ASST:
/
A
% foY h
r 1
i
'* k J
I
- 9. MARLESS INTERNAL DISTRIBUTICN
, ~ -.; eie.1i of AMT s15 dMS SITI SAFETY C I
A cec eve es M vnC onn o.
VAmnw m 'eCO F_av LION ANALYSIS i
i - ta e McFROFnrM arfinovi (DENTOM & '4ULLER Ir ea" fit icT* cut eT_D -
8>
a l net n
! IacsSICX 4 STAFF TNGINEE3 DIG l i
380f 'PJ l
l.
e 6 ssvarre,
- rvtmr*
r.
onsA i
IENYTRON TECH i IERNST i
- i.
y onevte I
c -uwe N t I ow-wt*TNG SPAC*0RS IBALLARD j l -wr I
e swr -r et i i c'et T n (vntSCELCOD i i 2nm ewh j g fI ee-c* vr AGevr';*
DrAC OR S At - IT SFA0 S*** TTCH j 1 2 -n g-nt,;
/ 9055 l 9AER IGAVMIT.L I 21 i i e, ccef-s
/l NOVAK l i FC-' EM f
~
I *** ANALYSIS i
f I ct'e M ROSZTOCL'Y I IGM"ES S
I"'Ik-M
/
CHECK l l I IVOLL'TER 1
t Iwe,--.<es i
I I i 19CNCM i
j
'c-Q AT C !
l I I
IJ. CO L*. Dis 3
i i i ! e At --vxt i I l 1: Sec. R i
~
- 3 i i or-- me n c i I i i i
EXTcRNAL CISTRIBUTICN l
CONTMCL NUMBER i
i i
6.n - *in.
8 8
! i 1779 I.-;
I I I I
'l y302A0203 l
e.
s--
I i-
' 8 e, se- - t;. wxtC :I--) 1 I I i l
ft 1.s Ys 3E'.r CA IOCR*f1 t#
TO AC2S I i il I
e I
..-..se.
- -e.
-