ML19296B448
| ML19296B448 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 01/31/1980 |
| From: | Ades M, Jensen S, Sofer G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19296B432 | List: |
| References | |
| XN-NF-79-089S01, XN-NF-79-89S1, NUDOCS 8002200577 | |
| Download: ML19296B448 (44) | |
Text
... _ _. _ _ _ _ _
XN NF-79 89 SUPP. 1 REYlSED LIMITING LARGE BREAK LOCA ANALYSES FOR FORT CALHOUN USING THE ENC WREM-IIA PWR ECCS EVALUATION MODEL JANUARY 1980 YYY$$N$0$
mmn
- . w
,. +,: n 2 --
~
- z:
=
1 RICHLAND, WA 99352
~-.n
+
_1[~
- ~
4 r
- +
ki<evs%sMsuknanM viwsNue
- ^ % w'r a
--u w
~
a
=m-n-
ERON NUCLEAR COMPANY,Inc.
j 8002200
e XN-f4F-79-89 Supp. 1 02/05/80 REVISED LIMITING LARGE BREAK LOCA ANALYSES FOR FORT CALHOUN USING THE ENC WREM-IIA PWR ECCS EVALUATION MODEL Prepared By:
S. E. Jensen 2
M. J. Ades Concur:
A%w O
J. N. Morgan,Panager Licensing and Safety Engineering 4
Approved:
/,Pf-gcy G. A. (aop 6 Map ger Nuclear Firels Engineering ERON NUCLEAR COMPANY,Inc.
I I
I I
I NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY I
This technical report was tierived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the UFNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilire Exxon Nuclear fabricated reloari fuel or other technical services provided by Exxon Nuclear for liaht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration W
of comoliance with the USNRC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:
A.
Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor.
3 mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; of B.
Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.
I I
g
_F.
I
I I
-i-XN-NF-79-89 Supp. 1
'g g
REVISED LIMITING LARGE BREAK LOCA ANALYSES FOR FORT CALHOL'N USING THE ENC WREM-IIA PWR ECCS EVALVATION MODEL I
TABLE OF CONTENTS
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 LIMITING BREAK ANALYSIS RESULTS 4
3.0 EXPOSED FUEL ANALYSIS RESULTS 30 36
4.0 CONCLUSION
S
5.0 REFERENCES
38
_LIS1 0F TABLES Page Table 1.1 Fort Calhoun Exposure Heatup Analyses Results for ENC 3
and CE Fuel 2.1 Fort Calhoun Limiting Large Break Event Times 7
l 1.0 DECLG BREAK.....................
LIST OF FIGURES Figure Page 2.1 RELAP4-EM Blowdown System hodalization for Fort Calhoun PWR 8
2.2 Axial Peaking Factor versus Fuel Rod Length for Fort Calhoun 9
ECCS Analysis 2.3 Blowdown System Pressure, 1.0 DECLG Break........
10 2.4 Blowdown Total Break Flow Rate, 1.0 DECLG Break 11 2.5 Pressurizer Surge Line Flow Rate, 1.0 DECLG Break 12 2.6 Single Intact Loop Accumulator Flow Rate, 1.0 DECLG Break
. 13 3
-ii-XN-NF-79-89 Supp. 1 LIST OF FIGURES (Continued)
Figure Page 2.7 Double Intact Loop Accumulator Flow Rate, 1.0 DECLG Break.
14 2.8 Average Channel Inlet Flow Rate,1.0 DECLG Break......
15 2.9 Average Channel Outlet Flow Rate,1.0 DECLG Break.....
16 2.10 Hot Channel Inlet Flow Rate,1.0 DECLG Break 17 2.11 Hot Channel Outlet Flow Rate,1.0 DECLG Break.......
18 b
2.12 PCT Node Blowdown Cladding Surface Temperature, 1.0 DECLG Break......................
19 2.13 PCT Node Blowdown Volumetric Average Fuel Temperature, 1.0 DECLG Break......................
20 2.14 PCT Nod Blowdown Heat Transfer Coefficient, 1.0 DECLG Break. 21 2.15 PCT Node Blowdown Depth of Zirconium - Water Reaction, 1.0 DECLG Break......................
22 2.16 Containment Pressure Versus Time, 1.0 DFCLG Break.....
23 2.17 Normalized Core Power,1.0 DECLG Break 24 2.18 Core Reflooding Rate,1.0 DECLG Break...........
25 2.19 Reflood System Pressure,1.0 DECLG Break.........
26 2.20 Reflood Downcomer Mixture Level,1.0 DECLG Break
.....27 2.21 Reflood Core Mixture Level,1.0 DECLG Break........
28 2.22 Cladding Surface Temperature During Heatup for ENC Fuel At BOL, 1. 0 DECLG B rea k..................
29 3.1 Cladding Surface Temperature During Heatup For CE Fuel At BOL 32 3.2 Cladding Surface Temperature During Heatup For ENC Fuel AT E0L..........................
33 3.3 Cladding Surface Temperature During Heatup For CE Fuel at 32,600 MWD /MTM Peak Pellet Burnup 34 3.4 Cladding Surface Temperature During Heatup For CE Fuel At E0L 35
- XN-NF-79-89 Supp. 1 1.0 Ir4TRODUCTION AND
SUMMARY
A complete LOCA-ECCS large break analysis was performed and reported previously for the Fort Calhoun Station operating at 1500 MWt (1). This document presents results cf a reanalysis of the limiting large break calculation which corrects system input data and analysis, and addresses the NRC concerns regarding flow blockage.
The analysis was performed using the ENC WREM-IIA PWR ECCS evaluation model (2.3,4). Final results are summarized in Table 1.1.
These results support an ECCS total peaking limit. F
, of 2.53 over the core life for both Exxon Nuclear Company.
Inc. (ENC) and Combustion Engineering (CEl fuel types.
~
The ECCS limiting fuel exposure cases are beginning-of-life and end-of-l i fe.
The results in Table 1.1 show that for the ENC fuel type, 10 CFR 50.46 criteria for calculated peak clad temperature and metal water reaction are cet in both cases when the 2.53 total peaking limit is used in the calculations.
The same is true for CE fuel for peak pellet exposures up to 32.600 MWD /MTM.
The ECCS calculations for the 42,400 MWD /MTM peak pellet burnup end-of-life exposure case for CE fuel were made with a total peaking limit of 2.46.
The calculatirns for this case again show conformance to 10 CFR 50.46 criteria.
The 2.46 total peaking in the ECCS calculations for CE fuel at end-of-life is 3% below the ECCS limit of 2.53.
This 3% reduction in ECCS allowable total peaking for CE fuel at end-of-life is considerably less than the reduction in actual operating total peaking that occurs because of fissile depletion for fuel assemblies that have been exposed beyond 32,600 MWD /MTM peak pellet burnup. Thus,in practice the achievable F for exposures above
m XN-NF-79-89 Supp. 1 32,600 MWD /MTM peak pellet burnup will be sufficiently below the achievable F for low exposure fuel, that the high exposure fuel with a 3% reduction in ECCS allowable F will not become limiting.
Therefore, a single F limit of 2.53 assures conformance with 10 CFR 50.46 criteria for both ENC and CE fuels to the maximum exposures calculated.
A reanalysis of axial peaking at the 90% core height was performed, and the results confirm that the previously calculated axial peaking relationship remains valid (Figure 1.1 of XN-NF-79-89).
~
M s
m
w - umur' -
m umms uns unus TABLE 1.1 FORT CALHOUN Exposure Heatup Analyses Results for ENC and CE Fuel ENC FUEL CF FUEL Exposure, PPBU (MWD /MTM BOL 48,000 (E0L)
BOL 32,600 42,400 (E0L)
Total Peaking, F 2.53 2.53 2.53 2.53 2.46 q
Peak Clad Temperature (PCT), F d80 2195 2012 2188 2190 Max. Local Zr/H O - Reaction, percent 4.6 9.1 6.2 9.5 10.1 2
Hot Rod Burst Time, sec 31.6 29.3 28.5 26.6 26.1 Hot Rod Burst Location, ft 7.47 7.47 7.47 7.47 7.47 Time of PCT, sec 208 252 229 235 254 PCT Location, ft 8.22 8.22 8.22 8.22 8.22 pg
%:b
=T Max, Zr/H O Reaction Location, ft 8.22 8.22 8.22 8.22 8.22 2
Linear Heat Generation Rate, kw/ft at BOCREC 0.8218 0.8682 0.8206 0.8596 0.8338 Total H Generation, % of total Zr reacted <1%
<1%
< 1%
< 1%
< 1%
2
y
-4 XN-NF-79-89 Supp. I 2.0 LIMITING BREAK ANALYSIS RESULTS Subsequent to the large break LOCA analysis performed and reported for Fort Calhoun, (1) comments were transmitted to ENC resulting from Omaha Public Power District's (0 PPD's) review of the system input data for the Fort Calhoun ECCS analysis.
These comments identified an error in the computation of the fluid volume contained in the lower downcomer, lower head, and lower plenum regions of the reactor vessel.
Correction of the volume data reduced tte volume in these regions by about 160 ft.
Further interrogation of the Fort Calhoun LOCA analysis results by ENC revealed that incorrect single channel flow rate data had been utilized in the T00DEE2 single rod heatup calculations. At the same time, the NRC raised concerns regarding the flow blockage models used for LOCA-ECCS analysis, and required a reassessment of flow blockage for each plant using new NRC methods for flow blockage. Therefore, a revised LOCA ECCS analysis was undertaken for Fort Calhoun to correct the above described deficiencies and address the NRC concerns.
The reanalysis was performed for the large double-ended cold leg guillotine break (1.0 DECLG) determined to be limiting from the previous break spectrum analysis. This break remains limiting with the required revisions since the effect of the revisions are small and since they apply uniformly with similar effects to other breaks. Hence, recalcu-lation of the previously identified limiting brea' with corrected input is sufficient to demonstrate conformance with 10 CFR 50.46 and 10 CFR 50 Appendix K for these small changes in results.
_ __ XN-NF-79-89 Supp. 1 In the recalculation of the 1.0 DECLG break with the identified changes, the system representation, all other input data, and the code versions used were identical to those used in the. previous analysis Limiting break event times from the reanalysis results for both ENC and CE fuels are given in Table 2.1.
The RELAP4-EM blowdown system nodalization from XN-NF-79-89 is shown as Figure 2.1, and the axial power distribution is shown as Figure 2.2.
Blowdown results from the revised calculation are shown in Figures 2.3 through 2.9.
Blowdown Hot Channel results are given in Figures 2.10 through 2.15.
The differences in blowdown results from the previous analysis are a direct result of the reduced system volume.
Containment pressure and e> '. ended normalized power are given in Figures 2.16 and 2.17.
The calculation for the containment pressure was revised to include the mass and energy released during both blowdown and reflood.
The reflood results which are obtained using the revised contain-ment pressure are given in Figures 2.18 through 2.21 and the T00DEE2 temperature results are given in Figure 2.22 for ENC fuel at beginning-of-life.
Final heatup results for both ENC and CE fuels are provided in Table 1.1.
Comparison of reanal/ sis results with the results of the previous analysis shows little difference between the thermal-hydraulic behavior.
The principal differences being a faster blowdown and a reduced refill time which result directly from the smaller system volume input to the analysis.
The correction of single-channel flow data and use of the
. XN-NF-79-89 Supp. 1 NRC blockage model affect only the final heatup portion of the transient.
A recalculation of axial profile sensitivity with a power paak assumed at the 90% core elevation having a peaking of 0.874 times the 70% core height value gave a PCT 18 F below the PCT for the 70% core height.
Thus, the axial peaking relationship given in Figure 1.1 of XN-NF-79-89 remains applicable for the revised analysis.
1 w
1
r
_ _ _ _ _ _ _ XN-NF-79-89 Supp. 1 TABLE 2.1 FORT CALHOUN LIMI"ING LARGE BREAK EVENT Th1F5 1.0 DECLG BPTAK Fin H.'O C5 FUEL Fi. _
BOL 12Uyg r.1sf TIME (seconds)
Initiate Break 0.05 0.05 Safety Injection Signal 0.55 0.55 Accumulator Injection.,
Intact Loop (s) 15.9 15.9 Pressurizer Empties 8.35 8.35 End-of-bypass 19.02 19.07 Safety Injection Flow, SIS 20.45 20.45 Start of Reflood 30.23 30.27 Peak Clad Temperature Reached 208.
229.
w M
.--i-----.
~-
M "-" M ' W W
M M
M M
FF E M
67 66 SECONDAW 64 c5
$5 SECONDARY 54 STEAM GtNERATOR STE AM GE NE RATOR g
e 22 g
LOCP 2 LOOP 1 15 la 31 30 PRE SSURIZ E R TUBES (
'21
,21 SG TUBES (
37 3?
25 I
UPPE R HE AD I4 II q
32 N
s 10 20
36
'16
. 24
, 40 26 INLET O UTLET 15
~ OUTLET INLET T
PLE NUM PLE NUM PLENUM PLE NUM CONTROL UPPE R 933 ROD PLE NUM 18 HOT LEG 19 35 34 HOT LEG SH ROUDS 12 13 18
~
33 28 27 25 48 41 45 11 9
OLD LEG M
M 44 COLD LEG (d2 37 y
17 14 22 Q3 UMP PUMP E
'50
'43 d47
'69131
'27 70 O
UPPER CORE U
8 Z
r 49 42
{G O
I7 al o
N "I
20 19 38
=
3J 6
23 8
41 40 34 35 51
]30 W SUCTION gg
[
E
.'3 10
- PUMP 5UCTION PUMP SUCTION 2* Si 3
h
,55 2
5 24 39
'9 O D ACCUMULATOR
=
EE ACCUMULATOR h:
sx
=
68 I
4 7
h k
>]
'8'
'12 2
COtD LEG COLD LEG COLD LEG 54 43 44 45 O
gx 163
'61 Z
SUPPORT ASSEMBLY jZ k
50 z=
h b
59 kT E
U 52
'60 32 g
56l LOWER PLENUM g
" 47 49 e
'58 51 CONTAINMENT 53 57 LOWER HEAD AC ~UMULATOR I
48 Fiaure 2.1 RELAF4-EM Blowdown System Nodaliza+. ion for Fort Calhoun PWR
~
e i
.i-i
...--iiy.-i------
- c' x2e210%esWc r
M (CVV.
e.
]$
0 u
N-6 l
t 0
I i
3 C
s is 6
?
y i
sl C
ua sn rA e
vS L
C
/ rC X
oE i
(
f t
C
, cn T au H F o G
h I
gl E
na H i C k
i f
5 E at 0 R er O P o C
F l
E ar V io I
xf T A E
i A
h L
I
- 4. L t
0 E ? g F
R n
C 2
e R
L P
e r d u o R
gR E
I J
4 i
F ~1 i
C C
e P
u F
0 x
O I
2 0
C H
LA C
I T
I C
F C
F L.
C
. ~
5 j
io aioc or<><
SO2U N cC 5 a. [
+
t -
- XN-NF-79-89 Supp. 1
.t'
-r-------r------7------------r------r------<-
1 l
l M
x to 8
co La
~'
_J uua Q
o 4
as L
m t
m I.I m
?
1
' lil I
u I_
N w
i
/
c
,/
~,ii y
/
8
'/
"ca
/
,/
g i M
.1 N
/
as
/
L 3
,/
.o
,/
u
/
/
w
/
7
/
1 i
i i
i i
ffP C0t!
COD 2 cr.11 00 0 006 00; T1 bdd l HISd ) '3dD% 3Ed HDN3ld 83ddD
- XN-NF-79-89
~
Supp. 1 g
i i
g a
~
b Qu CD O
J U
LAJ w,
c, o
e d
s b
42 fC
=
eN
^
b o
ld (o
a
- a"
$u
[d CD
- E H
5_=-.
F~
o c
4 3O V
3o O
_ g e
N eu 3
Cn
- Tr Q
m M
Y t
t g
g g
ca.
a cac a
a c
-e r l' t?35/Ul) 'rold >iU3du 16101
.. _... _ _ - - Xfi-NF-79-89 Supp. 1 3..--
.._---...--.7.~~~~'T 1~~~~
~~ T ~ '~~~~~'~~
l i
g v
g 6
e J
Uus a
~
~
?.]
o J
.e a
to
.(d o
O E
LD to
..a o
d g g E
cn H
5-p-
am L
u 0
'tam m
E o.
g m
d f
an m
.m u_
-r
^
y 3
005-COOL 00C 0002 0051 0001 009 9
' C FI' t '3S / Ol ) '91b3 M0ld 3NI l 39 DOS 832IMn%3dd
__ Xil-f1F-79-89 Supp. 1 i
- a a
m E
m
~
- )
Oaa LLJ Q
o N
y
.e 4J m
x 30 y
b u
o S
1d m
an 3
a O 'i b
~
~
t U
I
<t H
a O
Oa
'3 4J U
m
+>
C w
- d m
m W
~ ~
w 3
G I
I i
l l
I ccc cer:
eco:
com com ocs cet 9
.a Pi-( n?A / 61 ) 'J001 I 31SNIS-31bd M0ld Wn300
_ XN-IlF-79-89 Supp. 1
..,._._.r-------,-----
r-r-----------
a m
O Lm
~
l}
Oa U
LLJ Q
O.
a W
+J m
C%
1o n
C s_
O e
- u, m
n m
J
=E 3
s':a O
h E
e aoO a
+>
Um 4J C
M e
sq 5
3 0
Q N
a
~
Q 5.
3 C1 e
u.
I I
I t
t c
000.
0009 coes cect ctec ecor occ1 c
i C MI'
( 33S/81' 'JDO~1 I TlUn00-31bF M0ld WR110
h i Xfi-IIF-79-89 Supp. 1 t- -
i' i
i i
i C
I g
a m
- i E
m o
_a Owa
~
~
3 O
.e ae c:
~
- L 2
o u.
IU u
to o
1 O g, I
M c
c m
.c U
o C1 Eo><
- c..:.
A e
L s
en
~
C
~
iD i
i i
i i
00009 000Cf C000L 00002 0000" 0
0000T-000012 I r t
( 13C. / 81 ) ' Mold 131NI ~13NNUH3 9AO
.. _ _ _ Xft-ilF-79-89 Supp. 1 r
i i
i o
i l**
a
<o
- -j 2
cc eaou.a N
o J
c) e ec e
o 3
o
,3 u
'J p
n a>
-a
t o 5
Z H
"w I
C C
<a
_c u
as en to L.
O><
s.'
e N
Ok N
I u.
/
~
m i
00092 C0000 0005' 0000T 0005 0
co0r-00006 r
r M F' (33S/07) ' roll 1010C 13NNdH3 9AU
. - - Xft-f1F-79-89 Supp. 1
-- - -r i
T i
i i
g
.3 a
m 1-m O
J d
Q O
ll co gj' :<
e t.,i 3:O f.L M
k e, fr' Y
P.'
c i
w E
I m
- H j
l~
se O
I ej b
1
.?
u.
1 E
==%.
MM bW wwm OSI 001 04 0
05r-001-OSI-00-8 FMr-( 19S /W 61 ! MOl-1 131NI 13NNbHO 10H
^
..... _. _ _. _ _ _ XN-liF-79-89 Supp. 1 g
i i
i i
i i
3 l
x m
'yj Q
xm ea U
LAJ a
~
-4 O
e o
tu o
w x
g y 3
d S lu u.
m y
m o
- ae g
Id o i
ji c-d E
c Id 4c 1
- H
[.
4.Jox
~.
<o I
m O
L 2
Ch c
l
.i T
_ _ _ _ _. = -
_____m i
000 OST 07.
08 Ct c
Ot-0@
T i r l
( 13S /H dl i M0ld 131100 13NNbH3 10H i
__ Xil-flF-79-79 Supp. 1 t
i i
r-i i
- 3 l
ze L
/
m
/
e i
a
/
u l
2 8
o N
.e L
3 E
o a
LU E
(D y
N e
x v
a,,
eu
{#
3 m
m G (d
[
C f
N o
e C
G iu e
v 3
_ *A e
3 g
e M
z i--
oc f
~
m g
J
/
en E
I O
l l
occ oceT com con 0021 coo:
ces ec3 ST85 (d 930) 3Mnidd3dH31 336.d80S 0013
.... - - - - Xti-flF-79-89 Supp. 1 r
i i
i i-g I
a
/ -
~
g
,/
ea
/
y l
o
. o, 2
B 2
-4 eaE
('
0 g
tun 8
/
~
I u.
d fu
<a
()
L-g u.i n ii
' gi U
i E
/
it u
/
c e
/
c tu
,/
I O
mH >
/
l-e i
3
,/
8 6
g O
/
0=
HU n.
m
/
g C1
/
u-f t
t 00S2 0002 00R 0022 000?
CCtri 009T 00ff SI1S d 9301 3801083dH31 39083AU 3Hn10A
. _ _ _ _ _ XN-NF-79-89 Supp. 1 Trrrnvr --
,irrvr,
r r r r r r-..
, cwr -- - i mi a
n mCLm oa v
,:.1 ea a
o A
~
- z ce g
O C
l11 ln as o
O
-4 g
't' L
G S
til h
tr c
O 2
H
~
~
n ny
- i d 4
g t-(L G
5 10 6
I c
- H 3
F--
m W
"Uo
=
to go CL e.A A
~*
c e=
CD eu.
N u.__._
lu.LU '
' i v i i_1 t La i mi n i i t i i_xi CT 0T OT
,0T OT q
9 ETC i.3 ^30-214 -dH /ni d ) INkIDI-U300Il3JSNbdiib3H
i--sium
- - - - - Xit-Hd-l6-86 sndd I
e
- N 3C a
P S
-.ir o
8
.C 8
m
>t@
ot c:
~
'Y N, e
vm g
y m
8 e.., m.
a U
)1 I
D o
iF g
Z u
u; H m
O
=
e
.Q 0
,e w
tif h
m A
~
N-a
~
N E
i i
i i
820 0 420 0 020 0 610 0 210 0 800 0 400 0 0009 5
91MD JSLIM( NOITCAER RETAW-LATEM FO HTPED l
.mwI x= =7hoo 3
wc o*
k
,~
a er B
G L
n C
E D
0 1
a e
m i
T E
s R
e u
U 5
s s
C r
R e
F V
c r
e e E
s s r t
M i
,.. u N
E s I
M s O
I e
T T
N r
j P
,r e
o i
c.
t t
n e
E m
D n
i 0,
E ta n
N o
j O
C W
L O
C E 6
T u
1 R
O 2
F eru
,1 g
y 4
s n
s F
i Sf Es',nwzc-.Z'tZHSE;_o
__ XN-NF-79-89 Supp. 1 c
l r,
t
.I l
c t*.
g I
x I
mo L
co
~
k C
e 1
(!
I l
u LaJ a
s n.
a.
i
~
rt b
(
m 3
i.
o l.
C-C O
m L
c O
a'I U
I.
"O it O
a 5
to m
m
-.x 2
H o=
N M
~. 4 O
La G
.-u.
.I,,
a,,,
1 o"
1_o!
~En ON 33 mod 03ZI'ldW3DN
... - -.. - - - - XN-fiF-79-89 Supp. 1 o
J l
i I
I I
i
[$
tS 8
a m
W Lm g
a a
a u 8 w
G7 o
A gna o
3 o
n
.J u.
bJ c
P [r n
m x
8 o
E t
u_
e aw o
az b
s a
s m
r m
.4 Lo I
8
-4 g
1 1
9 l
I f
1 I
a s
t 2
1 9
??S/N1 '9108 SNI0007.4 3803 s
I
__ XN-ilF-79-89 Supp. 1 a
I I
I I
I g
a to O
L m
0 c3 v3
._J m
uw Q
u 10 O
07 o
- ' a g
a o
o e
"w
.-(L L
to o-
_ gm e
.o a
z u
o Q
m m
u o iu 8
az g
e f-g cn
^
o M
w
.4 N
e L
a og g
1 I
I i
1 Ir, st tt et at ac s2
,1
~'N*d UlSd '3EDSS3Md WGN3Id 83ddn
.---siium-
-e ii
- Xil 'lF-79-89 Supp. 1 r
r i--
i 1
i i
w>
n i
io
<>e a
me
- s u
m e
_.J u
LJ o
c.
o
.e,
..)
tu o
in
>e d
L o<> a e
k
(~-
2 r_.
y
.J X
- e s
\\d n tr L
n o
u d-O f' ?
U C
x
!t o
9
- GV v
<a-o
,H O
O
~
o' o
m m
C La en s
b) s 1
W i
2 i
1 n
c n'
n m
y 91 bH 1 -f
' l3 A TI iMn1XIW M3WDJNM00
---'M
imi-i-m i
__ XN-t4F-79-89 s
Supp. 1
.-.._.r-----
7 r-i
--r i
8 g
s E
\\.
~
s m
W L
cc ea gj g
o u
o
\\
LJ to w-
.1 e
- C o
e a
o
\\
.J e
\\
u.
L
=
ta c> tt x
r, r
=
\\
o a
u:
s-
\\
I L-O u
<>u u
g
~ az 8
\\
H F-
%e CE
~
1
~
s m
\\
e x
5-2
\\
c,
\\
s k"
.a k
N
\\
o Ns
\\
N N
l l
)
t.
c m
s 3
2 7
14 1i *llAT1 32nixIW 380J
ii
1 1
,\\De
>Z82TI J@sDo N
C t
o 7V.
e-LC i
l f
M{,iI
. l-]
kaer B
G L
C 1
i E
D 0
1 L
~
i O
.x B
5 N
t A
s s
N N'
l N
e i
1 F
u N
s C
N E
s' r
x' o
s N,
f I
N 1
p
~
u t
N t
a e
N a.
H
~
/
g n
E
\\'
1 I
i r
r m
r u
D
+_
e s
u
\\
T t
m a
r e
p me T
e c
7 i
1 v
a f
i T
F F
ru Z
'4 S
g T
E T
n G
D Q
f
/
1 0
i 4
4 t
d 7
t
/
d E
D D
c E
E a
D D
C
'd a
i C
s t
T
?
- 4 C
T t
P
(
/
C J
F R
2 2
T 2
\\
1 2
er l
u g
i F
~
3 ?,
,,oa
,S, 8*.
3)N j
4'
/
J r1a r_)$, We TtLf 9I)i.t UMaaGJU J
1
.-.-..-m_.---
. XN-NF-79-89 Supp. 1 3.0 EXPOSED FUEL ANALYSIS RESULTS This section presents the results of LOCA-ECCS analysis performed for exposed ENC and CE fuel.
The ECCS exposed fuel calculations for both ENC and CE fuels were performed using the RFLAP4-EM/ HOT CHANNEL (Version ENC 28F) and the T00DEE2/MAY79 codes.
The analysis models are essentially identical to those described and referenced previously for Fort Calhoun (1}.
Differences from the previous Fort Calhoun analysis include revision of system blowdown data (as discussed in Section 2.0), correction of T00DEE2 single channel flow data, and use of the proposed NRC Cladding Swelling and Flow Blnckage Model(5)In applying the NRC flow blockage model, fuel rod internal pressures corresponding to the ENC model (0) for nominal conditions were used.
The system parameters used for the exposure analysis are identical to those reported previously (I}
The core upper and lower plenum boun-dary conditions used for the BOL and exposed fuel HOT CHANNEL calculations are from the respective limiting break blowdown calculations, and the reflood rate versus time for the T00DEE2 calculations are from the limiting break reflood calculations, as reported in Section 2.0.
Table 1.1 provides the analysis results for ENC and CE fuel types at beginning-of-life (BOL). The corresponding T00DEE2 heatup transients for ENC and CE fuel are shown in Figures 2.22 and 3.1 respectively. The calculated PCT's are 1980 F for ENC fuel and 2012 F for CE fuel.
The maximum local metal-water reaction does not exceed 7'A for both ENC and CE fuel. These results are within the ECCS criteria limits of 2200 F and 17% and support an ECCS allowable total peaking of 2.53 at BOL.
m
..-.m.
___ XN-NF-79-89 Supp. 1 Results for exposed fuel and end-of-life (E0L) are also shown in Table 1.1 for both ENC and CE fuel types. As a result of increased stored energy due to burnup enhanced fission gas release and high flow blockage resulting from increased rod internal pressure due to fission gas release, the most limiting fuel exposure is at E0L.
The PCT for ENC fuel at E0L is 2195"F.
For CE fuel, the PCT is 2188 F at a peak pellet burnup of 32,600 MWD /HTM with a total peaking, F
, equal to T
?. 53. At EOL, the PCT for CE fuel is 2190 F and corresponds to an Fq value of 2.46.
The T00DEE2 heatup transients for ENC fuel at EOL and CE fuel at exposures of 32,600 MWD /MTM and at EOL are given in Figures 3.2, 3.3 and 3.4, respectively.
The maximum local metal-water results are less than 11% for both ENC and CE fuel types. These results are within the 10 CFR 50.46 limits so that an ECCS allowable total peaking of 2.53 is supported throughout the life of both the ENC and CE fuel types. As mentioned earlier, the CE fuel with exposure in excess of 32,600 MWD /MTM will not limit reactor operation since the reduced allowed F value 2.46 is much higher than the actual peaking that occurs when peak' pellet exposures exceed 32,600 MWD /MTM.
m
--mii---umum--i-um
-m
_ Xft-i1F-79-89 Supp. 1 4
. - ~. - - - -. - -..
r e
'J '
/
-J Ccc c-M 6
4 s
ts y
C 3
LL LAJ U
g O
/
a
/
3
'A p
/
r 1
m U
(5
/
- E
/
O C
II
'F f
(.)
.-~. '
/
o.
3' 1 t1 o
l e.1 1
w n
/
E N h
"m
, x
/
c
!d q) ti r.,
t I
a.
H E
,t-c; H
s' e,,
~
(o Q
1,1 y
a m
o g
3
<n cn a
.C t~
s r-
+)
3
+1 D
-e fj m
~
w,
u a
a a
\\
6 a
g.,
O.x w
,. o c.
= '
i? b' d
i n.
L
.)
Q 4
a u
/
3
.G y
Lu L
i i
i i
i i
et:
2-n ert
- or 2m
- x
- d 4 '27Te^30 - 3h!Dibd3dWii SNIC001
_ _ _ XN-NF-79-89 Supp. 1 7-.--._.,
- _7_ _-- _7 p
-aow t
~
n>
es 4
3
/
/
u.
/
/
/
o=w L
O a.
a
/
so c.
B i
/
M
/
m O
t I
.a P
1
/
t
/
O,
/
C l
- P CI g
/
L i.s 3
\\
/
o o
\\
/
v.
Q
\\
/
i n
s O
\\
ti u
s e
1 N
L x
0 N
W
(
~
s a
g i
e t-c' s2i as y
a m
v r,
+
I L
", l 3m c) a C
e I
- m N
m w
y m
PJ fj
=*
p==*
u
~
<1 m,
o a
ea
- d l-N o;
iT RM M
N m
}
}
O o
'\\
2 c
3
\\
3 L
t
/
/
u-
~
(
l I
l.
._.L_
_..L__
t i
l_ _.__ _ _ _1 00f; 22:
.:T
.m m
aov p;'-
nM 1 ' 73e'3C - 3enibbidWM ONICOU T
--,a_
y av4 x2.=m ~QaCW s
O g
mCuv* -
e i.
~
l;ll, 9
'c' y
N 1
N le 0
u 3
F z
E C
r g
o F
c p
t 8
u t
a y
e H
pu g n n r i u r B i
I m
u m
D t C
e E
S e
l O
r l
P N
u e U
O t P T
? C a
A I
- S e a E
r k E
i e
H 4
p e m P e
P E
T M M
T 4r I
e M T
c /
a D m
c i
e.
f W r M E
u D
S 0 0 O
g 6,
n i 2 d 3 d
N i
1 I
1 c
2 a tA l
0 T
T n
F F
C u
2 7
C 2
4 C
i 7
3 T
T E
i P
Q 0 A c
3 O
a t
F i
4 1
L t
e C
0 O E c E r
C O
a 0 u
N C u 0 N
N g
t
(
?
I i
C u
F P
a 2
i 1
z I
~
2 I:
n 4.
S
- ?>.
, l'
- n eua
- 3 s
zWi (zHoOuJ._"
- J 1 ooWmJwa
- wXa"
r m c v un.. a n c. _.
ec z.,= ;._
j i
i i
a:- cae i N 'DE ts :.i 2. :t L T. s SI
[
e vieta
=c f
1 cot u a-
.. n,
/
N
/
x
/
,, k i
a ti J
t n"
I m
2
/
i a
M l
W a
w' 6 <;
I 6-i J
9' s
o 3 ~'
^
,\\
l x
x i
wx 2
C 3
s u
a.
~.
u=
l
=
'l E
>-* N e.
s m l
1 t
I i
i i
t i
i
,~,
u 4
sv x~
.: ~
c-u: -
TIME - accNa; Figure 3.4 Cladding Surface Temperature During Heatup For CE Fuel At E0L
__ XN-NF-79-89 Supp. 1 4.0 CONCLilSIONS The reanalysis of the limiting break for Fort Calhoun shows that with the appropriate corrections to the Fort Calhoun ECCS analysis the emergency core cooling system will continue to meet the NRC acceptance criteria as presented in 10 CFR 50.46 with the Cycle 6 core for both ENC and CE fuel and future ENC reloads of similar design.
That is:
1.
The calculated peak fuel element clad temperature does not U
exceed the 2200 F limit.
2.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of Zircaloy in the reactor.
3.
The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The hot fuel rod cladding oxidation limits of 17^4 are not exceeded during or after quenching.
4.
The system long term cooling capabilities provided for previous cores remains applicable for ENC fuel.
These acceptance criteria are satisfied if the Fort Calhoun reactor is operated at 1500 MW(t) within the maximum linear heat generation rate (LHGR) of 15.22 kw/ft, the F of 2.53, and the axial profile limits computed previously as in Figure 1.1 of Xii-llF-79-89.
Operation within the allowed LHGR limit and the allowed axial offset limits will neutronically preclude total peaking above the 70% core height level from reaching the limits shown by Figure 1.1 of XN-NF-79-89.
__ XN-NF-79-89 Supp. 1 Operation within the LHGR limit and F of 2.53 will also assure that the high exposure CE fuel will be unable to approach the reduced F limits from Table 1.1 reouired for this fuel at high exposure.
--mens- - - - - - -- ----------
m
. XN-NF-79-89 Supp. 1 5.0 _ REFERENCES 1.
Exxon Nuclear Company, Fort Calhoun LOCA Analyses at 1500 MWT Using ENC WREM-IIA PWR ECCS Evaluation Model, XN-NF-79-89, Septer.ber 1979.
2.
Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Mode, XN-75-41:
a.
Volume I, July 1975 b.
Volume II, August 1975 Volume III, Revision 2, August 1975 c.
d.
Supplement 1, August 1975 e.
Supplement 2, August 1975 f.
Supplement 3. August 1975 g.
Supplement 4, August 1975 h.
Supplement 5, Revision 5, October 1975 i.
Supplement 6, October 1975 j.
Supplement 7, November 1975 3.
Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-II, XN-76-27(A),
XN-76-27, Supp.1(A). XN-76-27 Supp. 2(A), March 1977.
4.
Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA. XN-NF-78-30(A),
Xil-NF-78-30. Amend. 1( A), May 1979.
5.
D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," Draft NUREG-0630. November 8,1979.
6.
Exxon Nuclear Company, Flow Blockage and Exposure Sensitivity Study for ENC D. C. Cook Unit 1 Reload Fuel Using the ENC
~
WREM-II Model, XN-76-51. Supplement 1. January 1977, XN-NF-76-51 Supplement 2. January 1978; XN-NF-76-51, Supplement 3, March 1978
-m
Xft-NF-79-89 Supp. 1 02/05/80 REVISED LIMITING LARGE BREAK LOCA ANALYSIS FOR FORT CALHOUN USING THE ENC WREM-IIA PNR ECCS EVALUATION MODEL Distribution MJ Ades DJ Braun RE Collingham GC Cooke RD Hyman SE Jensen WV Kayser DC Kolesar JE Krajicek TL Krysinski DC Lehfeldt CD May JN Morgan LA Nielsen GF Owsley GA Sofer PJ Valentine Document Control (10)
Omaha Public Power District (20)/
C. D. May lluclear Regulatory Commission (40)/
G. F. Owsley
- -. -. - - - -