ML19260B610
Text
TD *. DftfMLL EtwNh:d f Wi FKor 4 E omley sogou uucuum commu,ino.
2101 Horn Rapids Road F. O. Box 130, Richtend. Wsahington 99352 Phone (509) 9438FCC Tehur: 32-6353 November 4,1979 Mr. Darrell G. Eisenhut, Acting Director 27M Division of Operating Reactors Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission Washington, D. C. 20555
Dear Mr. Cisenhut:
In my letter to you dated November 2,1979 the Exxon Nuclear Co@any (ENC) reported the results of its review of the recent rod burst data presented by the NRC staff at the November 1,1979 meeting.
This review concluded that the new data was bounded by the presently NRC approved ENC rupture and blockage model.
This comprised the basis for the initial finding that ENC's ECCS analyses satisfied 10 CFR 50 Appendix K(I.B.)
The staff further requested that ENC evaluate the impact of the NRC Staff's proposed rupture / blockage r.odel.
In response to this request ENC has perfomed exposure analyses using the NRC Staff ructure/ blockage model in the ENC Evaluation Model.
The plant chosen was considered to be the most sensitive to the NRC rupture / blockage relationships because of its slow temperature ramp rate in the LOCA analysis.
Cases were run which encompassed fuel life (0 to 38,000 MWD /MTU peak pellet exposure).
The current Fq limits for the selected plant were used in this analysis.
The results of this analysis were tnat the Peak Clad Temperature (PCT) remained below the 22000F. limit for all exposums.
The rupture stress varied from 2.6 at BOL to 3.7 Kpsi at 38,000 ffWD/MTU in these analyses.
On the basis of the above, ENC concludes that there is no adverse impact on ENC licensing calculational results from the new rupture burst data or use of HRC rupture / blockage model.
Furthemore, plants licensed with ENC evaluation models continue to be in confomance with 10 CFR 50.46.
Sin rely, D*"}D
~9'}
o Ju S. k o
d d c Ju G. F. Owsley, Manager Reload Fuel Licensing Exxon Nuclear Company GF0:pbp 7912100456 1519 071 AM AFFILLATE OF EXXON CORPORATION
Enciosure 3 3.0 -
3 4
\\
+
+
'39 30 36 2.0 -
b g,,.
11 12 9
1.0 -
14 15 16 1,3 20 26 21 16 a3.
Io o
o'.5 1'.0 i.5 2:0 25 X (IN.)
Computer simulation of B-1 section of 76.5-cm elevation showing maximum and minimum flow restriction definitions.
1519 072
4 e
8 et
\\
m m
\\
o ce
~$
bo
\\
\\
~
^
A
-$ e cu N
\\
N)
M
\\
8 M
N cnD
/
-l @
c4 s4 m
m b
/
m l
O
.$ mm n
/
E-xo
\\
O
\\
M
-8s it
- h IYUI 7VI NSHg NDO 1519 073
M & m>W37 => Kt @ oOX' OC k mw 00 1
)
(
0 A8 E
RA WOa L
F l
g I
l N
N
/
I N
N O
\\
,/
IT N
/
C
/
U D
ER o
g o
o 5
n e
t m
E$m~$ moo 5 hmCM mMTb Q
l 6
D I
1 V
S20 6l9l REDUCTION IN FLOW AREA (%)
"P
?
8P 8P 25 o
a
////77D/h M a.
f l
5
\\
x
\\
w b g_
\\
f o
m f \\\\\\TNNNA i
l-
!/
i a
/
5 9
0 a
=
o c-g m
- = g
!,l$E a<5 8 g
8 E
8 F
B&
3:
W 2
5 (T
+
)
F B
A kn&
S T
T E+
M
-R P,
j' I A
Eu f
RA M
P T.
UR B
E UR f,,
(-
S D
T E
G.
S C.
)
/
A IN CUR V
E S
e O
4 e
s x
3 3
\\
x
\\\\\\ ti i
r, e
l l
ND~%)
h 1
-I 6
l g
T e
o>
g e
(%)
IVEIS 'IV 13 P002 Ogggg 1519 077
e 6
C-E (PRO.) & FAST-RAMP BLOCK. CURVES a
5 d 8~
Q sb f
c t
5
\\
o g_ s
/
tn s
7 e
N
/
3 g_ N g
/
\\
s c-s
/
g
\\_
a N\\
/-
h h
s o
ENGINEERI G HOOP STRESS (KPSI)
~a Fig.O
GE & SLOW-RAMP BURST STRAIN CURVES
~
UPPER GE CURVE POR INSIDE RODS M
LOWER GE CURVE FOR OUTSIDE RODS a
z 8-
<m H(n8-th A
-o 4
E S-z N
CG No A*_
N g,_
A__
GE o
__a o
i i
i i
i 000 700 800 900 qqo 1000 1100 1200 u,
TEMPERATURE (DEG. C) e CD
- h. h Ne
e D
ENC (PWR) & FAST-RAMP BLOCK. CURVES S
S
< 8-
/[
/
Q x
/
N \\\\
d
- g_
O N
/
on d
\\
h
\\
/
me /
t s
2 I
~
Z
.gN O
o_
//
=
a a-a a
3 ENGINEERING HOOP STRESS (KPSI)
E, Fg.cm3
e 52 SEVRU C
0 )I 2 SP
.K K
C
(
O S
L S
B ER P
5 T 1
M S
A P
R O
OH W
N O
G L
b N
l S
I
[N /
R E
-,'[7x E
)
N R
I W
GN P
5 E
(
C N
N E
\\
2 6
4N x
x 1
~ ;y F
0 8
g R!$
.eh:*
d 1
g_*S_
9 8
I' I
I
_s*
i i
t
\\.!eg x
s.m i-h W B(
d
\\
o h
~k b
2 0
.i,
.i
.l PA) NIVEIS 'IYLUEHEBULJ1113 ben $N1I$I 1519 082
~
e$-
gt tNzs mh<% 8 r
1 s
_o E
8 W
P JO W
70 f
0
/
C OM T
W
~
P E8 O
0 M0 P g/ x
\\
S IT E/
RO E
AT B
U9 U
0 R0 R
E w
ST (9
D4
~
S E
G. 6/'/
T f
1 R
00 A
0
/
C I
)
N CU 1
1 R
0 V
0 ES 12 J
00
fi ESeE Zr< E
- S$!4 o
a 8
8
~_o 0,
7 3:
)
5 b
NN NN\\K\\
W E
0 N
/
G I
F N
A E
S b
T
//((7 E
I R
N 1, 0
G AM H
. I' O
w P
OP BL S
O T
1, C
5 R
>s K.
ES x
S C
\\
U (K
R P
V S
E 2
I) 0, S
25
\\
0
980 6tSt
- S ') ~
REDUCTION IN FLOW AREA (%)
0 20 40 60 80 100 g
t e
i t
r
't
/
M o_
\\
4 N
s m
m N
l F
to O
3 l
4
$ 5-l 5
E O
T m
I m
A
% s_
o
/
O
/
?
/
nC t
W m
m 0 0~
td U2 3
Enclosed are 3 figures that show correlations in the 10/31/79 draft for rupture temperature, rupture strain, and assembly flow blockage.
Temperature ramp rates are accounted for in the correlations, and the ramp rates that are most appropriate should be used.
If it is not practical to accomodate ramp rates in the code, envelopes of these curves should be used. The tabular values from which these curves were generated are also enclosed.
~1 s
i519 086
b
-R O 8
o
.8 a:
-g N 8-
~Y I
-e h s
O esc ca, U 'M )
g CR 27pg 1519 087
e 1
1 1
m 8
1 i.
1 t
\\
.}Q I
f
/
8.
5 y
J..
lo 7.-
V
./
/
1
/
-g 1
1
\\
s ~
-p
~
~
\\*
e
- TM wIxa*IMS
(.3
,x.
wo a
=
3s n e,88
0 I
8 1
n E
I f
lnn k
-g o ggl
=
i C
l l
C
-S $
a j
03 m
/
/
8 m
/
/'
O
~&
y
/
a g
l i
m s
-e$e
~
m s
N
~~'"........3.-
g,
. ~............,.,. \\,
e e
a s
O an m
W OP W
G M ysay AOW NI NOUDDGHH bbN ONO$0$d[
1519 089
Slow-Ramp Correlations O'C/S
<10*C/S
<10"C/S Burst Engineering
~ Burst
-' Flow Temperature Hooo Stress Strain Blockage
('C)
(KPSI)
(%)
(%)
1 600 15.56 20 15.2 2
625 14.40 21 17.6 3
650 13.22 30 26.1 4
675 12.03 44 39.4 5
700 10.84 58 54.6 6
725 9.68 70 66.5 7
750 8.53 78 72.7 8
775 7.40 80 74.6 9
800 6.30 79 73.6 10 825 5.24 72 68.4 11 850 4.26 54 50.8 12 875 3.36 39 34.7 13 900 2.59 31 26.6 14 925 1.98 30 26.1 15 950 1.52 31 26.6 16 975 1.20 29 25.2 17 1000 0.97 22 18.5 18 1025 0.80 20 10.2 19 1050 0.68 20 15.2 20 1075 0.59 20 15.2 21 1100 0.51 20 15.2 22 1125 0.45 20 15.2 23 1150 0.41 20 15.2 24 1175 0.37 20 15.2 25 1200 0.33 20 15.2 1519 090
Fast-Ramp Correlations 28*C/S
>E5 C/S
>25 C/S Burst Engineering
~ Burst Flow Temoerature Hoop Stress Strain Blockage
(*C)
(KPSI)
(%)
(%)
1 600 31.14 20 15.2 2
625 28.74 21 17.6 3
650 26.39 23 20.0 4
675 24.01 26 22.3 5
700 21.65 33 29.0 6
725 19.32 50 47.5 7
750 17.04 63 60.3 8
775 14.78 67 64.6 9
800 12.57 67 64.6 10 825 10.46 63 60.3 11 850 8.50 54 50.8 12 875 6.70 39 34.7 13 900 5.17 23 20.0 14 925 3.95 22 18.5 15 950 3.04 23 20.0 16 97E 2.40 34 29.9 17 1000 1.94 57 53.7 18 1025 1.61 78 72.7 19 1050 1.36 80 74.6 20 1075 1.17 78 72.7 21 1100 1.03 37 32.8 22 1125 0.91 35 30.9 23 1150 0.81 37 32.8 24 1175 0.73 47 43.2 25 1200 0.67 57 53.7 1519 091
Composite Correlations Burst Burst Engineering Flow Temoerature Strain Hoop Stress Blockage
(* C)
(%)
(KPSI)
(%)
1 600 20 31.14 22.3 2
625 21 28.74 22.3 3
650 30 26.39 22.3 4
675 44 24.01 22.3 5
700 58 21.65 29.0 6
725 70 19.32 47.5 7
750 78 17.04 60.3 8
775 80 9
800 79 14.78 64.6 10 825 72 7.40 74.6 11 850 54 6.30 73.6 12 875 39 5.24 68.4 13 900 31 4.26 50.8 14 925 30 3.36 34.7 15 950 31 2.59 26.6 16 975 34 2.40 29.9 17 1000 57 1.94 53.7 18 1025 78 1.61 72.7 19 1050 80 1.36 74.6 20 1075 78 1.17 72.7 21 1100 37 1.03 32.8 22 1125 35 0.91 30.9 23 1150 37 0.81 32.8 24 1175 47 0.73 43.2 25 1200 57 0.67 53.7 1519 092
DRAFT 10/31/79 D. Powers / R. Meyer CLADDING SWELLING AND RUPTURE MODELS FOR LOCA ANALYSIS D. A. Powers and R. O. Meyer 1519 093
1.
INTRODUCTION During a postulated loss-of-coolant accident (LOCA), the reactor coolant oressure may drop below the internal fuel rod gas pressure causing the fuel Core cladding to swell (balloon) and, under some conditions, rupture.
behavior during a LOCA would depend on the time at which swelling and rupture occurred, the magnitude of swelling, and resulting coolant flow blockage (i.e., reduction in flow area).
Such phenomena were among the many reactor safety issues discussed during the 1973 rule-making hearing on Acceptance Criteria for Emergency Core Cooling Systems (ECCS). The adopted accepta..s criteria (Ref.1) limited predicted (calculated) reactor performance such that if certain oxidation and temperature limits were not exceeded, then core cooling would be assured. It was required that each licensee use a safety evaluation model to analytically demonstrate compliance with the acceptance criteria.
Appendix K (Ref. 2) gives requirements for some features of evaluation models, and, in particular, states that to be acceptable the swelling and rupture calculations shall be based on applicable data in such a way that the degree of swelling and incidence of rupture are not under-estimated. The degree of swelling and incidence of rupture are then used to calculated other core variables including gap conductance, cladding temperature, oxidation, embrittlement, and hydrogen generation.
After the conclusion of the ECCS hearing, the AEC reviewed and approved cladding behavior models for each U.S. fuel manufacturer for their use in ECCS analyses.
1519 094
During the ECCS hearing uncertainties were apparent in predicting fuel behavior during a LOCA. Therefore, in the Commission's concluding opinion (Ref. 3), the Comission directed the AEC's research office (now the NRC Office of Nuclear Regulatory Research) to undertake a major confirmatory research program on cladding behavior under LOCA conditions. The resulting multi-million dollar program includes simple bench-type Zircaloy tests, single-and multi-rod burst tests that simulate some in-reactor conditions, and actual in-reactor tests ranging to full-size bundle tests.
The research programs are not all finished, but with the completion of many out-of-pile and a few in-pile tests, we are at a plateau of under-standing that greatly exceeds our understanding in 1974, and the results have not confimed all of our previous conclusions. The trend of these recent data shows the likelihood of more ruptures, larger rupture strains, and greater flow blockages, than we previously believed.
Consequently, we see the need to reevaluate all LOCA cladding models to assure that licensing analyses are performed in accordance with Appendix K.
In the following sections we will display the relevant body of data, describe our evaluation of these data to arrive at useable correlations (curves), and compare these correlations with those currently used in licensing analyses. Since the data show strong heating-rate
- effects,
- Both heating rate and strain rate' are important tactors in detemining cladding burst pressure and strain. However, most burst experiments are r;ot designed to distinguish between heating-rate effects and strain-rate effects. For the purposes of this report, the actual differences are probably unimportant. Therefore to avoid confusion, in the remainder of this report we will refer to both effects simply as heating-rate effects. 1519 095
we have derived different curves for slow ramp rates and fast ramp rates. But most current ECCS models do not include a ramp rate effect, so we have also displayed composite curves that envelope the slow-ramo and fast-ramp curves. 1519 096
2.
DATA BASE The ballooning and rupture behavior of Zircaloy are fairly complex phenomena in part because (a) the stresses are biaxial and the material is anisotropic in the temperature range of most interest, (b) the properties of zirconium-base alloys are susceptible to heating-rate effects, (c) oxygen embrittlement increases yield and failure strengths, and (d) the cracking of oxide coatings results in failure sites that can localize stressas. Consequently the behavior of Zircaloy depends strongly on the cladding's environment and hence on test conditions (Refs x-y).
Therefore, for final calibration of the data correlations, we have selected only those data from experiments in aqueous atmospheres that utilized either internal fuel-pellet simulators (i.e., indirect cladding heaters) or actual fuel pellets in reactor. This selection emphasizes the more recent and more expensive proto+"nical test data and deemphasizes much of the earlier data. Appendix A provides a tabulation of all of the data we have used, their references, and a legend of symbols that are used for these selected data sets in the later figures.
There are holes in this data base, however, particularly with regard to the absence of large bundle tests, and we have utilized the results from simpler less typical tests to bridge the gaps. These more pristine tests are atypical in a sense, but they do reveal fundamental features of Zircaloy behavior that allow one to interpret the sparser prototypical data. 1519 097
3.
NEW CORRELATIONS 3.1 Rupture Temperature The incidence of rupture depends on the differential pressure across the cladding wall, the cladding temperature, and on the length of time those conditions are maintained. Time duration under burst conditions manifests its&lf as a heating-ramp-rate effect, and this effect will be treated exp1tcitiv, Wir have converted differential pressures to hoop stresses to eliminate design-specific dimensional effects. The conversion was made using the thin-shell formula, o = (d/2t)aP, where o is cladding hoop stress, d is the undeformed cladding mid-wall diameter, t is the undeformed cladding thickness, and AP is differential pressure across the cladding wall at rupture. Table 1 shows some computed values of hoop stress in terms of differential pr9ssure for comon commercial fuel designs.
Figure 1 shows rupture temperature dr.ta as a function of hoop stress for a wide range of test conditions. While this figure shows the general trend -- tubes bJrst at lower temperature when the pressure differential is higher -- the data are scattered primarily because of ramo-rate effects and experimental uncertainities in detemining burst temperature. 1519 098
VENboR F0Et MooP STRESE
(.?sr)
too DPSI 2co4Psr Iceo )Psr 2 coo bPsr MW 45 xis ITx i C-E 14 Xs4 85355 16 x it.
W i+xs+
15 Xs5
("I%l'I f6 7 Y~1 txe 2xe R ERC c E sys+
cfanir W IW89 v/ iSUS q C 7 V "7 QE 858 QE t*th y[Q hkkkke
_ c__
1519 099
N C3y BURST TEMPERATURE VS STRESS DATA g_
mO 56 b
Miso
' ' ~
- x g,
Y es k
o px i g.
g.
e-x
- s.
O *t)fji A 7
a; CD e
s i
0 6
B B
M ENGINEERING HOOP STRESS (KPSI)
L.
n-
Figure 2 shows ORNL data at 28*C/sec (a connon ramp rate used in th? ORhl experiments) and the basic correlation we will adoot as developed by Chapman (Ref. Q) using numerical regression techniques.
It is clear that most of the data scatter has been eliminated by restricting the data to a single ramp rate. Chapman has also developed a ramp-rate correlation (Ref. N) that can be used with the basic rupture-temperature correlation in Fig. 2 to produce a family of rupture-temperature curves. Ramp-rate has little effect on rupture temperature for rates faster than 28*C/sec.
Three curves that span the important ramp-rate range are shown in Fig. 3 along with the data of Fig. 1.
Chapman has shown that most of the original scatter is explained by ramp-rate effects, and the curves in Fig. 3 are seen to span most of the data.
The up-facing triangles still deviate from the correlations and the major body of data.
Difficulties in temperature measurement for these TREAT in-reactor data (Ref. X) are believed to be responsible for th's deviation, and such discrepancies will be seen in later disolays as well.
3.2 Burst Strain Defonnation (burst strain) at the location of a rupture depends on temperature, differential pressure (which is related to temperature by the correlation in Fig. 3), ramo rate, and several other variables such as local temperature variations. These effects have been discussed previously (Refs. x-y).
Figure 4 shows burst strain as a 1519 101
~
m Bmu 4BpMm oma c8 a 8_
2 o8 m
~
8
/
M C
)
/
I S
BU E
R l
x N
S G
T I
N T
E E
E M
R I
P N
E G
RA H
T O
U OP RE ST C
R U
E R
S V
S E
(K P
2 S
8 O
I c
)
/
R s
NL DATA o
O
a AT X
AD S/
e T,
C e )S I
T 8
~
A 8 8 P
/C K
(
4 x1 S
S S
E E
V s
R R
/c aT U
V o
S C
P E
OO R
H U
g T
G A
eN I
R OO R
E E
P E
M e
N E
I T
G N
O T
aE SRUB A
o g
g U
e
@@g em$g-3 G-
bo
- ~t_.
n.
9?NaeE6'5EGT M
s,3e s
3 J
S) J3 Oe%OD%NN MZh4e1-WE
~Z $-
s a
o O
- o 8
w8 Q
6 y' r 00
,S i
BU 7
x R
00 S
x T
A
+
S v
4 +
+
. E 2
g
- g.9g T
o
+ g +,
- o v
3 T
R y
t E8 y
A 0
a M0 h
I P
x* $ a N
x
.0+.
+
E x
+
R V
g v
A
- 44 x+
S T
x g'
+
U0 d
T 0
0 R0 g
a O
E x
E M
(
P D
E a
E R
G1 A
0 0
T 0
C
- x U
)
R n
x E
x x
x D
1 x
A 10i 0
TA x
x 1200
function of one of these variables, burst temperature, and the data scatter is therefore due to temperature measurement dif-ficulties and the other variables mentioned above.
The scatter in Fig. 4 is bewildering., so we have relied on data from less prototypical but more controlled tests to help derive a correlation. Figure 5 shows burst strain versus burst temperature from Chung and Kassner's work (Ref. T) with short
'Zircaloy tubes heated by passing an electrical current directly through the Zircaloy. Several fundamental features are apparent.
There are three superplastic peaks -- one in the low temperature alpha phase around 800 C and two in the high-temperature beta phase around 1050'C and 1225'C. The very important valley at about 925'C is a consequence of mixed alpha-plus-beta-phase material, wnich exhibits low ductility.
Heating-rate effects are also visable; slow-ramp rates produce large strains in'the temoerature regime below about 950'C as a result of feedback effects discussed in Refs. x-y.
But slow-ramp rates produce very small strains at temperatures greater than about 950't because the Zircaloy has time to oxidize and embrittle before significant ballooning can occur.
Fast-ramp rates produce the opposite effects in both temperature regimes.
To derive the slow-ramp correlation, which is shown in Fig. 6, we have thus taken Chung and Kassner's 5 C/sec curve and scaled the peaks and valleys to pass through the more orototypical data in our data base. The alpha-phase peak at 775 C was assigned the value C
-l'.-
1519 105
l.8 CLADDING CONSTRAINED WITH MANDREL
^ '^
^
1.6
- HEATING RATE 5 C/s H
I*4 O HEATING RATE 55 C/s N
o HEATING RATE 115 C/s P
l.2 e9 5
/ *'s h
- 1. 0 0
f S
2 of a
i S
0.8
- [
I BURST IN STEAM g
a As i
- 4 2
0.6 s' *t 8
i*
.os
/
g 0.4 4
q 7-*
4
\\
o O.2 o
s n.
o 600 7UO 8vc 900 1000 l100 1200 1300 BURST TEMPERATURE ( C) o c
f
-l.5 -
519 106 u
003 1
oo n
)
0C 0
'0 2G (D
E E
oR dU a T M
A T
R E
P ol d h aE T
'+,*+h-o d
y' 00
=
8 R
s B
g e
R 8m U
g mg a G
of 80% in order to bound Chapman's 10'C/sec bundle test. The five highest points in Fig. 6 (0-10'C/sec heated-shroud single-rod tests) are preliminary and have not been fully evaluated, but they were disregarded because the heater power was so low (about 3W) that the tubes were in effect burst in a muffle furnace (the heated shrouds). Direct or external heating methods are known to exaggerate rupture strains by maintaining artificially small local temperature variations (see Ref. X), and such experiments were excluded from our data base. Since tha atajority of the data is bounded by the curve, we believe that the correlation satisfies the intention of Appendix K noc to underestimate the degree of swelling.
It should be cautioned that some very recent, unevaluated data from Germany (Ref. X) also show large strains (up to 120%), so the potential exists that Fig. 6 may have to be revised upward.
The fast-ramp correlatioa is shown in Fig. 7.
In this case, there are no data from prototypical bundle tests and limited single-rod tests with heated shrouds and uniform heaters in the area of the low-temperature peak. The correlation was obtained by scaling Chung and Kassner's 55'C/sec curve and adjusting the alpha-phase peak height in relation to the peak height in Fig. 6 according to the relation that a 28 C/sec peak would have (based on interpolation) in Chung and Kassner's curve (Fig. 5) to the 5 C/sec oeak in Fig.
5.*
When prototypical bundle tests and heated-shroud tests are performed in the future, we expect the data to fall near the curve in Fig. 7.
- Consideration is being given to adjusting these curve peak locations to higher temperatures -- to around 825'C. 1519 108
O I
6 4
x a
.xx y
m i
e
-gl.
=
-al ix a
.a s.*-
-gh m
a
- x b
5 4
x i
d s
.I J
W W
W 09 OF W
0
(%) NIVH.LS WNID f$$$ $N1Nb g g 109
Figure 8 shows the composite (i.e., envelope) of the curves in Figs. 6 and 7 along with all of the data from Fig. 4.
The composite curve gives a good representation of the data, providing that the causes of small strains (Ref. X) are kept in mind.
3.3 Assembly Flow Blockage Very few measurements of bundle blockage have been made under prototypical conditions and the best attempts are shown in Fig. 9.
It is therefore necessary to derive bundle blockage from single-rod M
burst strains, but this is not straight forward test results have shown that ruptures in a bundle are not coplanar.
Figure 10 is a cross section from Chapman's first bundle test (Ref.
X). Notice that only a few of the rods have burst in this plane.
We have chosen the most realistic (minimum flow restriction) of Chapman's definitions of blockage for the following analysis.
Figure 11 shows the axial distribution of blockage for Bundle No.1, from which the maximum blockage is seen to be 49%.
Figure 12 shows the geometric relation between average rod strain and bundle blockage for a square array of commercial-size tubes.
From this figure it can be seen that an average rod strain of 27%
would cause a bundle blockag ef 49%. Since the average ruptura strain for rods in Bundle No. 1 was 42% (see Appendix A), the blockage can be obtained from the rupture strain by multiplying by 0.64 (the ratio of 27 to 42) and utilizing Fig.12. The similar ratios for Bundles No. 2 and No. 3 are 0.67 and 0.70 giving an overall average for the three bundle tests of 0.67. 1519 110
8
-1 s4 I:1
-8 x
xx x
sg
=
l g?
H n
l
. '.. d
- 4l
'dk~
l
~
q 4
,. <*a, t
f E
s N*s 3
~ '
A A
e e
e 6
(z) arvats wumasananza bbbbbO$0$d1 1519 111
O 9
LOCAL FLOW BLOCKAGE VS STRESS DATA 0
8 g s-m4
$ 8-e
+
- g_
zo b
x 0
D S-Q l
c O
i e
i 0
5 10 15 20 25
~
ENGINEERING HOOP STRESS (KPSI)
~
N 1,. y
ORNL-DWG 78-11137 20-l 77 MAXIMUM FLOW REsmCTION DEFINITION
}
MINIMUM FLOW RESTRICTION
[ DEFINITION
/
2.5-3
/
+
2 l
3 1
35
+
/
+
+
l l
s.
S. t /.
e 35 gg 11 5
$ 1.5-9 10 11 i
12
(
16 gg 13 14 II 15 is
+
+
1 20
+
4 as.
0
~
0 0.5 1.0 1.5 IO 2'5 IO X (IN.)
Fig.10 computer simulation of B-1 section of 76.5-cm elevatt.on showing maximum and minimum flow restriction definitions.
-QC -
15M \\\\5
100 5
e
% 70
=c s.0 i
a soi m
J\\
s i.0
/
V
\\
,/T 3
v x
e
/\\
\\l
\\
20 M
\\
(
O
-10 0
10 20 30 40 50.
60 70 80 90 100 DISTANCE FROM BOTTOM OF HEATED ZONE (cm)
Fig.ll. Coolant channel flow area restriction as a function of elevation in Chapman's Bundle No.1.
-ll -
1519 114
N
-R g
-E m4
-s su ER
~
a2>
-8 5
-S $
4 o fE W
- g
-R S m
-8 =e 5
4 S
-a 8 2.
-2 >4 a4
-S
-s
-a n
G a
a e
a a
4 4
4 4
aar a as oc as as at as as or o
(20 V H H Y X CPId NI N O' Q G G H E f_
k o k ) k /A\\L
-g.
1519 1i5
Assuming that the distributions of ruptures in Chapman's bundle tests are typical, the local blockage correlation is thus formed by multiplying strains in Figs. 6 and 7 by 0.67 and then utilizing Fi g. 12. We have called this result " local blockage," as distinct from the desired assembly blockage, because it does not yet represent large commercial-size bundles or include the effects of non-fueled tubes, which would not balloon. The slow-and fast-ramp local blockage curves are shown in Figs.13 and 14 where they are compared with the sparse collection of data. Figure 15 shows the composite flow blockage curve, which envelopes the curves in Figs.13 and 14.
Finally, to obtain assembly flow blockage, two adjustments are required. First, it must be recognized that bundle-average block-age, which is desired, is a function of bundle size. This can be seen by~ envisioning an 8x8 test bundle that is analyzed quadrant by quadrant. If each 4x4 quadrant is viewed as a small bundle, the planes of maximum blockage for the quadrants would be expected to occur at different elevations because of some randomness of the One would therefore expect to find the plane of maximum process.
blockage in each quadrant to have greater flow restriction than the plane of maximum blockage in the bundle taken as a whole. That is, the large bundle size introduces an averaging effect.
To account for this effect for commercial fuel bundles ranging from 7x7 (BWR) to 17x17 (PWR), we have used an average blockage from Chapman's bundle tests rather than the maximum value used in developing Figs.13 - 15 (that process was appropriate for the
,t' 1519 116
e d
1 R
4 l
-g'o
..I b
-S f g
s
.. H
=
g
>I C
d
==
6 e
A e
6*
~4 (M vaay mou N1 Nouonasu
(([$ $$1Nkm
+
gg m
O P
FAST-RAMP LOCAL PLOW BLOCKAGE de DATA g
B E
m
=
8 ma a-mE 3-
's
+
a Hg i
G E
R- '
~
u O-i i
0 5
10 15 80 35 ENGINEERING HOOP STRESS (KPSI)
Ill
d 8
d a
4 CI 4
-g o EE E
d E
-a g
s 8
m S
e
-S s m
N g
5
..g N
8 6
6 g,
6
~
(%) V:inIV KCTI4 NI NOLIOGGl int
$$.$$$INkI i
data comparisons because the bundles represented in Fics.13-15 were all small arrays). For Bundle No.1, the average (41%) of the blockages was found between the 23-cm and 47-cm locations in an attempt to eliminate the suppressing effect of spacer grids at 10 cm and 65 cm. Similar averages were found for Bundles No. 2 and No. 3.
Using these values the ratio to be used to derive large-bundle blockages from rupture strain data is 0.55 (compared with 0.67 for small arrays). This factor was used to derive all of the blockage curves in the next section of this report.
The second adjustment is a reduction of about 5% to account for instrument tubes and guidetubes that would not balloon. The exact scaling factor SF depends on the fuel design and is given by SF = N A /(N A + N A ),
rr pr gg where N is the number of fuel rods, A is the flow area around an r
p undeformed fuel rod, N is the number of guidetubes or instrument g
tubes, and A is the flow crea around an undeformed guidetube or g
instrument tube. This scaling factor was also emoloyed in deriving the blockage curves in the next section.
.. Q.
1519 120
APPENDIX A FUEL CLADDING BURST DATA DATA REFERENCE A (Upright Triang'.e)
FRF-1 R. A. Lorenz, D. O. Hobson, and G. W. Parker, " Final Report on the First Fuel Rod Failure Transient Test of a Zircaloy-Clad Fuel Rod Cluster in TREAT," Oak Ridge National Laboratory Report, ORNL-4635, March 1971.
Available in public technical libraries. Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.
R. A. Lorenz, D. O Hobson, and G. W. Parker, " Fuel Rod Failure Under Loss-of-Coolant Conditions in TREAT," Nuclear Technology, II, p. 502 (August 1971).
Available in public technical libraries.
Inpile, 7-rod bundle, steam atmosphere.
Maximum reduction in bundle flow area = 48 %.
Mean rod burst strain = 36 %.
Mean rod burst temperature - 389 C.
Mean rod engineering burst stress 1.71 Kpsi.
t R00 RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
( C/S)
(PSIG)
( C)
(%)
(KPSI)
H 25-36 172 966 26 1.39 4-1 25-36 250 799 35 2.02 R
25-36 205 743 36 1.66 4-2 25-36 290 816 42 2.34 L
25-36 162 915 36 1.31 I
25-36 190 827 35 1.54 C
25-26 215 810 40 1.74 1519 121
DATA REFERENCE B (Cross)
R. H. Chapman, "Multirod Burst Test Program Prcgress Report for April-June 1977," Oak Ridge National Laboratory Report, ORNL/NUREG/TM-135, June 1977.
Available in public technical libraries. Also available from National Tech-nical Information Service (NTIS), Springfield, Virginia 22161.
R. H. Chapman, J. L. Crowley, A. W. Longest, and E. G. Sewell, " Effects of Creep Time and Heating Rate on Deformation of Zircaloy-4 Tubes Test in Steam with Internal Heaters," Dak Ridge National Laboratory Report, NUREG/CR-0343:
ORNL/NUREG/TM-245, October 1978.
Available in public technical libraries.
Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.
Out-of pile, single rod, steam atmosphere.
R0D RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
( C/S)
(PSIG)
( C)
(%)
(KPSI)
PS-1 28 922 893 18 7.47 PS-3 28 809 873 29 6.56 PS-4 28 850 871 21 6.88 PS-5 20 830 882 26 6.72 PS-10 28 870 901 20 7.05 PS-12 28 891 898 18 7.21 PS-14 28 844 883 25 6.84 PS-15 28 893 885 17 7.24 PS-17 28 1760 778 25 14.2 SR-1 28 116 1166 26 0.94 SR-2 28 146 1082 44 1.19 SR-3 28 249 1011 43 2.02 SR-4 28 650 921 17 5.26 SR-5 28 1380 810 26 11.2 SR-7 28 2090 736 20 17.0 SR-8 28 178 1020 43 1.44 SR-13 28 155 1079 79 1.26 SR-15 28 2780 714 14 22.5 SR-17 28 154 1049 53 1.25 SR-19 28 2760 688 16 22.4 SR-20 28 154 1049 55 1.25 SR-21 28 162 1023 48 1.32 SR-22 28 129 1081 50 1.05 SR-23 28 139 1077 35 1.13 SR-24 28 144 1057 67 1.16 SR-25 28 139 1092 78 1.13 SR-26 28 120 1130 34 0.98 SR-27 28 133 1084 41 1.08 SR-28 28 1220 835 27 9.87 SR-29 28 1170 843 27 9.45 SR-37 28 1967 760 23 15.9 SR-38 28 1998 770 20 16.2 1519 122
DATA REFERENCE C (Plus)
MRBT-B-1 R. H. Chapman, "Multirod Burst Test Program Progress Report for July-December 1977," Oak Ridge National Laboratory Report, NUREG/CR-0103:
ORNL/NUREG/TM-200, June 1978. Available in public technical libraries.
Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.
R. H. Chapman, " Preliminary Multirod Burst Test Program Results and Implications of Interest to Reactor Safety Evaluation," paper presented at the 6th NRC Water Reactor Safety Research Information Meeting, Gaithersburg, MD., November 7, 1978. Available in PDR for inspection and copying for a fee.
Out-of pile, 16-rod bundle, steam atmosphere.
Maximum reduction in bundle flow area = 49 %.
Mean rod burst strain = 42 %.
Mean rod strain in plane of maximum blockage = 27 %.
Mean rod burst temperature = gag r, g(og Mean rod engineering burst stress = 8.72 Kpsi.
R00 RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
( C/S)
(PSIG)
( C)
(%)
(KPSI) 1 29 1124 852 36 9.10 2
29 1075 8 67 32 8.71 3
29 4
29 1052 860 36 9.33 5
29 1005 872 45 8.14 6
29 1104 872 43 8.94 7
29 1052 869 36 8.52 8
29 1074 872 42 8.70 9
29 1030 870 47 8.34 10 29 1059 873 45 8.58 11 29 1054 847 53 8.54 12 29 1114 863 37 9.02 13 29 1091 878 59 8.84 14 29 1066 875 42 8.63 15 29 1062 865 42 8.60 16 29 1092 848 39 8.85 i519 i23
DATA REFERENCE C (Plus)
MRBT-B-2 R. H. Chapman, "Multirod Burst Test Program Progress Report for July-December 1977," Oak Ridge National Laboratory Report, NUREG/CR-0103: ORNL/NUREG/TM-200, June 1978.
Available in public technical libraries.
Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.
R. H. Chapman, "Multirod Burst Test Program Progress Report for July-December 1978," Oak Ridge National Laboratory Report, NUREG/CR-0655: ORNL/NUREG/TM-297, June 1979. Available in public technical libraries.
Also available from National Technical Information Service (NTIS), Springfield, Virginia 22161.
Out of pile,16-rod bundle, steam atmosphere.
Maximum reduction in bundle flow area = 53 %.
Mean rod burst strain = 42 %.
Mean rod strain in plane of maximum blockage = 28 %.
Mean rod burst temperature = 858*C.
Mean rod engineering burst stress = 8.88 Kpsi.
ROD RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
(*C/S)
(PSIG)
( C)
(%)
(KPSI) 1 29 1117 870 35 9.05 2
29 1115 846 39 9.02 3
29 1096 853 40 8.88 4
29 1100 872 42 8.91 5
29 1127 866 35 9.13 6
29 1004 857 58 8.13 7
29 1067 861 56 8.64 8
29 1097 856 38 8.89 9
29 10 29 1065 856 43 8.63 11 29 1112 853 40 9.01 12 29 1094 851 40 8.86 13 29 1134 883 41 9.19 14 29 1048 858 42 8.49 15 29 1152 836 35 9.33 16 29 1117 848 42 9.05 1519 124
DATA REFERENCE C (Plus)
MRBT-B-3 R. H. Chapman, " Preliminary Multirod Burst Test Program Results and Implications of Int.erest to Reactor Safety Evaluation," paper presented at the 6th NRC Water Reactor Safety Research Information Meet.ing, Gaithersburg, MD., November 7, 1978.
Available in PDR for inspection and copying for a fee.
R. H. Chapman, "Multirod Burst Test Program Progress Report for April-Juae, 1979," Oak Ridge National Laboratory Report, NUREG/CR-1023:
ORNL/NUREG/TM-351, in publication.
Out-of pile, 16-rod bundle, steam atmosphere.
Maximum reduction in bundle flow area = 75 %.
Mean rod burst strain = 57 %.
Mean rod strain in plane of maximum blockage = 40 %.
Mean rod burst temperature = 764 C.
Mean rod engineering burst stress - 11.07 Kpsi.
ROD RAMP PRESSURE BURST BURST ENGINEER: %
RATE AT BURST TEMPERATURE STRAIN BURST STRESS
( C/S)
(PSIG)
( C)
(%)
(KPSI) 1 10 1393 771 48 11.28 2
10 1280 779 76 10.39 3
10 4
10 1318 767 55 10.68 5
10 1375 764 63 11.14 6
10 1327 770 51 10.75 7
10 8
10 1320 756 78 10.69 9
10 1320 754 59 10.69 10 10 1362 774 50 11.03 11 10 1396 775 57 11.31 12 10 1414 761 47 11.45 13 10 1486 760 49 12.04 14 10 1405 769 42 11.38 15 10 1335 753 53 10.81 16 10 1407 747 59 11.40 1519 125
DATA REFEREM E D (Closed Circle)
F. Erbacher, H. J. Neitzel, and K. Wiehr, " Interaction Between Thermohydrauli and Fuel Clad Ballooning in a LOCA, Results of REBEKA Multirod Burst Tests with Floodin;," paper presented at the 6th NRC Water Reactor Safety Research Informati:n Meeting, Gaithersburg, MD, November 7,1978.
Available in file for USNRC Report, NUREG-0536.
F. Erbacher, H. J. Neitzel, M. Reimann, and K. Wiehr, " Fuel Rod Behavior in the Refilling and Reflooding Phase of a LOCA-Burst Test with Indirectly Heated Fuel Rod Simulators," paper presented at the NRC Zircaloy Cladding Review Group Meeting, Idaho Falls, May 23, 1977.
Available in file for USNRC Report, NUREG-0536.
K. Wiehr and H. Sc;nnidt, "Out-of-Pile Experiments on Ballooning of Zircaloy Fuel Rod Claddings Test Results with Shortened Fuel Rod Simulators,"
Kernforschungszentrum Karlsruhe Report, KfK 2345, October 1977.
Available in file for USNRC Report, NUREG-0536.
F. Erbacher, H. J. Neitzel, M. Reimann, and K. Wiehr, "Out of-Pile Experiments on Ballooning in Zircaloy Fuel Rod Claddings in the Low Pressure Phase of a loss-of-Coolant Accident," Proceedings of Specialists' Meeting on the Behavior of Water Reactor Fuel Elements Under Accident Conditions, Spatind, Norway, September 13-16, 1976.
Available in public technical libraries.
F. Erbacher, H. J. Neitzel, and K. Wiehr, " Studies on Zircaloy Fuel Clad Ballooning in a LOCA, Resu'ts of Burst Tests with Indirectly Heated Fuel Rod Simulators," paper presented at the ASTM 4th International Conference on Zirconium in the Nuclear Industry, Stratford-o.a-Avon, England, June 27-29, 1978. Available from ASTM.
Out-of pile, single rod, air and steam atmosphere.
R0D RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
( C/S)
(PSIG)
( C)
(%)
(KPSI)
?
11
?
880 27
?
?
11 856 883 51 5.91
?
11
?
865 33
?
?
11
?
860 44
?
?
11
?
840 32
?
?
11
?
840 36
?
?
11
?
840 43
?
?
11
?
840 54
?
?
11
?
830 47
?
?
11
?
825 27
?
?
11
?
825 33
?
18 11 1420 823 33 9.81
?
11
?
820 28
?
?
11
?
820 38
?
14 11 1420 810 38 9.81 15/9 126
DATA REFERENCE D (Continued)
ROC RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
( C/S)
(PSIG)
( C)
(%)
(KPSI)
?
11
?
810 42
?
?
11
?
810 44
?
35 11 1380 794 27 9.54
?
11
?
780 27
?
?
11
?
780 30
?
?
11
?
780 52
?
?
11
?
770 26
?
?
11
?
770 32
?
?
11
?
760 24
?
?
11
?
755 23
?
?
11
?
755 52
?
1519 127
DATA REFERENCE E (0 pen Circle)
E. Karb, "In-Pile Experiments in the FR-2 DK-LOOP on Fuel Rod Behavior During a LOCA," paper presented at the US/FRG Workshop on Fuel Rod Behavior, Karlsruhe, June 1978.
Available in file for USNRC Report, NUREG-0536.
E. H. Karb, "Results of the FR-2 Nuclear Tests on the Behavior of Zircaloy Clad Fuel Rods," paper presented at the 6th NRC Water Reactor Safety Research Infor-mation Meeting, Gaithersburg, MD, November 7, 1978.
Available in file for USNRC Report, NUREG-0536.
E. H. Karb, "Results of FR-2 In-Pile Tests on LWR Fuel Rod Behavior," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Behavior Information Exchange, Idaho Falls, Idaho, June 22-29, 1979.
Available in PDR for inspection and copying for a fee.
Inpile, single rod, steam atmosphere.
ROD RAMP PRESSUR" BURST BURST ENGINEERING RATE AT BURL.
TEMPERATURE STRAIN BURST STRESS
( C/S)
(PSIG)
( C)
(%)
(KPSI)
A1.1 7.1 725 810 64 5.01 A2.1 20 1276 820 36 8.82 Bl. 6 8.2 1160 825 38 8.02 B3.1 10 1146 825 37 7.92 Bl. 3 12.7 885 845 34 6.12 A2.2 12.1 841 860 56 5.81 Bl.1 17.5 754 900 30 5.21 Bl. 5 9
653 910 60 4.51 Bl. 2 8.7 653 915 25 4.51 B3.2 12.1 725 915 50 5.01 1519 128
DATA REFEREN'CE F (Square)
R. H. Chapman, J. L. Crowley, A. W. Longest, and E. G. Sewell, " Effects of Creep Time and Heating Rate on Deformation of Zircaloy-4 Tubes Tested in Steam with Internal Heaters," Oak Ridge National Laboratory Report, NUREG/CR-0343:
ORNL/NUREG/TM-245, October 1978. Available in public technical libraries.
Also available from National Technical Information Service (NTIS), Springfield, Vi rginia 22161.
Out-of pile, single rod, steam atmosphere.
R0D RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
( C/S)
_{PSIG)
( C)
(%)
(KPSI)
SR-33 0
8?5 762 23 6.68 SR-34 0
844 766 32 6.84 SR-35 0
648 775 29 5.25 SR-36 0
660 821 29 5.35 SR-43 4
1105 773 29 8.95 SR-44 5
1060 777 30 8.59 SR-41 9
1416 757 27 11.5 SR-4?
10 1373 761 28 11.1 i5i9 129
DATA REFERENCE G (Asterisk)
REBEKA-1,
-2, -3 F. Erbacher, H. J. Neitzel, and K. Wiehr, " Interaction Between Thermohydraulic and Fuel Clad Ballooning in a LOCA, Results of REBEKA Multirod Burst Tests with Flooding," paper presented at the 6th NRC Water Reactor Safety Research Information Meeting, Gaithersburg, MD., November 7, 1978.
Available in file for USNRC "eport, NUREG-0536.
K. Wiehr, "Results of REBEKA Test 3," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Behavior Information Exchange, Idaho Falls, Idaho, June 22-29, 1979.
Available in PDR for inspection and copying for a fee.
Out of pile, 9-rod bundles, steam and water atmosphere.
TEST INITIAL MEAN MEAN MEAN MEAN REDUCTION RAMP PRESSURE BURST BURST ENGINEERING IN FLOW RATE AT BURST TEMPERATURE STRAIN BURST STRESS AREA
( C/S)
(PSIG)
(*C)
(%)
(KPSI)
(%)
1 7
870 815 29 6.01 25 2
7 800 870 53 5.53 60 3
7 725 830 44 5.05 52
) 5)ll
""U
DATA REFERENCE H (Inverted Triangle)
M. Bocek, "FABIOLA," paper presented at the 4th JAERI-FRG-NRC Annual Fuel Behavior Information Exchange, Idaho Falls, Idaho, June 22-29, 1979.
Available in PDR for inspection and copying for a fee.
Out-of pile, single rod, steam atmosphere.
ROD RAMP PRESSURE BURST BURST ENGINEERING RATE AT BURST TEMPERATURE STRAIN BURST STRESS
_ %)
(KPSI)
(
( C/S)
(PSIG)
( C) 1 3
563 860 66 3.92 4
11 1375 790 8
9.58 8
7.8 1375 780 35 9.58 10 10 2013 750 33 14.03 12 9
563 890 29 3.92 13 10 1810 765 10 12.62 1519 131
DATA REFERENCE I (Diamond)
J. L. Crowley (ORNL), personal communication to D. A. Powers (USNRC),
August 10, 1979.
R. H. Chapman (ORNL), personal communication to D. A. Powers (USNRC),
September 11, 1979.
Out of pile, single rod, heated shroud, steam atmosphere.
R0D RAMP PRESSURE BURST MAXIMUM ENGINEERING RATE AT BURST TEMPERATURE R00 STRAIN BURST STRESS
(*C/S)
(PSIG)
( C)
(%)
(KPSI)
SR-47 10 1436 775 MMyr"?S 12.35 SR-49 5
1139 775 98 9.80 SR-51 0
1030 790 93 8.86
- R-53 0
841 760 83 7.23 R-57 0
725 775 110 6.23 i519 132
EE_NfBAL APPENDIX K REQUIREMENTS MUST BE MET.
REVISED MODELS MAY BE REQUIRED FOR ALL VENDORS.
ALL BREAK SIZES NEED TO BE CONSIDERED.
IF UNCERTAINTIES ARE NOT CONSIDERED IN SWELLING AND R CURVES, APPROPRIATE SENSITIVITY STUDIES MUST BE PERFORMED.
WIDTH OF THE VALLEY MAY BE AS IMPORTANT AS HEIGHT OF T 1519 133
DETERMINING MAGNITUDE NEED TO SORT OUT SUBSTANTIVE CONDITIONS.
SOME TEMPERATURES AND RAMP RATES MAY NOT BE EXPE THEREFORE MODELS NEED ONLY APPLY WHERE CONDIT OCCUR.
PWR RUPTURE TEMPERATURES 840 *C - 960 *C BWR RUPTURE TEMPERATURES 960 C - 1200 C
1519 134
3 h
S b
ll l5
..g=e D.
I
/
E Ei
%i
/
6/-af
=
a f
b,
/
8*
n i
U
.o 1
/
~5 g
Y h
S i
s e
? -c h x
~
+
+
+
(a ma) ganwssamL 1519 135 E23 [2 Ild,
a 4
\\-._
_. --.I I
I i
. #.. i.
4 y,.
. i i..
~
I i i i e i 4
. e i
j.. t a
g I
f f. 1
, f f i,, *
. 1 f I f.
8.
. 4 a i a 6 e 4 I
1 9
e e g g j
l i n g
i i s 6 i i i
e a i..
. ~.
,l
. l l
6 t I a i f.
t s f I f
I e....
t t
. e
]1[
i
.. g 3 9 i
. i
, e i
1 5 6 6 i -t i.
i.. i i
. e l t e
, e 1 3 I t
ee.... s
.ae.
t i
i g.
,,... )
... i i.
i
, i
,, st..
g,...,
,, i,,
2, 4
t!i 4
i
,,, s,,.
. i i.i v6,,
i.
- x. x. g. i g
g, m,
u a..
> > i 4
i wi.i y
ii.a v iii n...
i i a i
,,... i,,,
ei i i
..x 1 1 1 ic f
.I i f f I e. t i I f f i 1.
s i
i\\'
1 4
6 6 1 #
i
. i i
m
<a a a e i # i i s i i..: ii Thh i
! '.1 f
m_.a.1,.u t _ i i i i i i, i ie
-u i
vi
./
i....
4 i
ii I,
c3 i i
e e
j a
a 3 1
t-i i,
f._,.a i '
I 3
1 I I
i e
,f 1
1 e e i
J 7.i4.
l '
8-r i i.'
i i 4
.y1,.ye n_ /,,
.o i
. i a e i
.. e I 4
i. <
t
. Oc. e.
- C.'.,'
e
.e
~. i e
8 8 I i i r i i i I i
u i
t i,,
. I i i% i I i ! I l
s/
i
.i i f f Il i i e a
g if l
I i.
I l
%u i t t 4
cy-}
[.,
/
I j.,
y
^%
- f. l I
I
. i a n
1 4
i i
p, e t l t i
4 4
I i i I e1 i t i i i,
t t i t i # t i i n i i e
i 1
I
,1 i
I
... i a i i e
- I e i i i
i I I i 6 '
6 i, i 1
i
! e i 4 Y
t y' '.
t, f -
t 3
l 6.
l
.g g
,,,jg,,,
e
.. e e e s...
s t
]ji
.. f........e
_g fM i e t :
f
[
T.
1 l
i.
p i
.. e i.
t,
s,
4 i
y f,
- a t
. I 4 4 5
y 1
L I e
e
. l t i
. t.
.p c i. s
...... i.
.. i,,
i a
. eJ [ g t e g~T-}
i i,, i
(
i i i
i,
, ' i '-.
N.
1 i i f f
t 9
_ i i
.t i
n 9 l,.
i
( l t i i
, t t
l
! i I r e s 8 t i
4.
f w t t
. g t
i g
, i t t
.i.
i 6.. i i i i j,
l l
W 6 E4 e
i i
..,,,,,, a r
s.
i 4 i
, 6 i
i a
,.. d_
I t I
It i. I t i 1
.s e
...I f 1 1,.f i 1 1 1 1 9 i e t,
~
7f f l t. 6 6
7 ~
1 e
~,.
i
, /.
/
- - _._.._.3 Q
W*
./
.i O&
CD I
.. _. n 4
i
.,X h
~
i _.
F*
r x i.
3 A
kI i 8 '
.r.
q.
f '
A I
f N
a
\\
if t
1 S
,,x.,
x i
.I i f f f
i t
, f f i ! '
_ 1 i \\
l I
i i
i
.. i.....,,,
1.,,,,,
a u..
a a
a
~
__ o
__m O
- _ o 4- -
o.
o c_
i w_.
2'
. nr
_._ _.._ N32h16
.illMR o ~ q[) _
-.. f.
...1 v
s 4 4
. ;.. i,1.3,1, g L1 i, i., gj g t 3_q t ! # # #
i t
i i r.
~
r i _
I a i!.LL 1, 1
'~
re i
i.
.m...
.t 8
'. 4
(
.~..
ppn o wy 5
Ju AJ
=
1519 136
4 9
I I
I 3_ _
_'f E
n M p
I t '. i
. t
.I t
r,
a f
..e 3
i l l..
I
.ll l[# l $ [.l-l t.
-. u
- -e r
.-r.
i i ;
[ T T..
iie i... e i.
,... 6 1
! t t t t t t i i I t.
. 1 i
1 1 e a i i 1 1 i
i 1 3.
1 i l I I 1
4 i 4 4 3.. 6. l l i sii.4 4 s t.
Q l
l 1 1 A. ggt
.. e...
. t,
e. i.
.1
. t t t t
'.. 6. s i e a i I i.
I I
I i..
. t,
t p.
g s
g g q.,
... e e - e i t t t i et t, e
- 8. 3.. 8.
- I r
, i e i i I f.
I.
6 I i
i t
t t 9 i
i i r i
i I
. 4 i i i i i
i 4
. 1 [ t i
i ;. y -
e r
i i..,
t e
l i I.,y Q
.. t i
I I
e
,e r
_ 1 !
I i i l
4 I
6.Q4 4
{
g g l, j g
[
~
i 3 4 4
1 i i 8 8 I I
p.
1 t
t
, n O
Lf
,g i
i e
i i
i e.+ii u
i i; i.,; ;
i i a i i e i e
i i,
%p (4
_. L i
. 3 6 I
i.
. P 7 L1,4 w-t'p i
i T-"'1 s i J i i sa si i i ei4. s s, #.
i.
- , g,F@
4 1
i i i i i i
. i i,
a, i
, i i
. i.
e t.
t i I !
., i
_n j
g l
t i t
I i
i 3
f
)]h.
1
]Mi f.
i i l i
t i i i t t.
i
_ *Q i
U r
i J/
i e N-Q e i i, i
+ 8 1. m.
u i l i i i i e i >>
W. \\,,#
6 o i 4
. i i..
1 6
1 i !
s'h e
t i i I e i e i
l i.'b 3
3 '
I I e i.
. fit M I 1 111e i t t t i.
I i
8 s i I 6 i i i s a i I 4 i i
,,. 7,
- .h4
- i p[ f
. 3 3 1 6
I 6 1 4 6 4 3 4 l
I I 4 i
. l.
6 0 $ g.t
.U,g
{ e 1
, t i
t t t i
I t
i e i
I 1
1 i I I I I it i e a i a
- A.e i e t i s i 1
e.
i i i I 1 i I I
I I i f f I I t. k.J t 1
4 Vt
' I i i
- g.
1 l
/
t i.
! N.
t i t t t t t i #
i
- t. t
' i g
i i.
4 4 8
f4 4
1 6 i n4 l e i I.
_J_
i i 3 4 l l l l l l
.;ff f
pg L t t. t i. I i e it i
. I i i
g L., _}__jt n, t e i
l 11
_e t
, 6. 4 I i!
I t
i I
6 T
.... i,i i
e i i i..
I i I.Le i 3 i i i 1 i
i I
t i
. r..
j t t t
. 1 (i H
! l.
. 6 6. 4 4 6 i 1 ei6. eie e i g
i i
g W
9.,'.
. s f
J 1 f.
... re
. i i i 1
i
, i, i 3
Ee 4
i s i i i,,, t....
i,
t r i g g
~Q
, t g g],
g g i
. 3 5 6 l 1 i g ) l
{ g l 4 g l g g g 9 g g g a
b.
i eai i n,. er e i i w.
t i
6 9
. e i 6
gn. git i
t.
l j j 3
. t #
1.
8
,f i.
1 et i.
i i
e i 6. i.
6
., t i
i.
i e
i, i i i i i i
i,,
, t
... i.U..
i. n-i 4
a i, i i. i i
r i e t
i i I8 t t t t ti t i r. i. i.
i..t
.. I t i e.
t i e
i
_r, i
u --
.i O3
.., i,
i i t
f f.
i n
~
i
/
r.n 3,.f tn.
i i_
..., i, d
\\I t'
i
.. i i,
,, t,.
2 i i i.....i-
. 4 Ni 4
i N..i i
t #
.N-i I
1 e i,. t i
i i i, i 4
i
.. i l
i.
t.
1 v
N.,
. i,,
.. i i.
v.
u.
.T - 4
_b 1
.x a i
3 i
iI
., i
.... # 4 N
\\
1...,,
6y
. i t i i i t i
.e i
I i
i k
l a
. 1 t
. I t
) $ l l l l 3 [ 3 I t t i 1. t I I l t ! I t I i i 1 i t i 1..
t I
t
! I i
, i,,...,
,.,.,,i, 1,
\\
_O 9
i I
i i
O o
6
- o,
_Q
-Q
. i i i.. i i
a
'r as o.
_d
_ t.
L_ - i. t i,,
"C. L.T.
e G':s 2,.2 UHdLLLL I I (J 4 (J L L.i i L.f~I.E_L t * '
D 8
- 1 i
i e i e6*3 i eaie 6e e e
t6
.t. g. :
a.
i i i i,..,,..,,.,
s
. 6 I
3.
I a
. I I
l I f
a f
g_Lm.2,..
_g
. I I 8 I e.
. I 12 d.li-..
- ..: J...
i e_
it e
1 I i t t t 3
{
t t i i e.
_.,..L..
r s.
t l.LLt _.
J.L t._.1. ' J. t.Lf.1 ' L J. l. l L a. t.
{
.. ?.!.. '
I ' '.
i J
- . L.L.. _ t LuL i u ; ;. -
- .
. _ L._ _....... _
'-' ' ~
~- '-
.T'~
'-', i Q PG fm f3 1519 l37 W
p w
IMPORTANT PAPAMETERS CLADDING TEMPERATURE FUEL TEMPERATURE BURNUP (FISSION GAS)
DIMENSIONS PLENUM TEMPERATURE POWER PLASTIC STRAIN HEAT TRANSFER v
v PIN PRESSURE (STRESS)
RAMP RATE v
RUPTURE TEMPERATURE l' STRAIN
~
BLOCKAGE e
CLADDING DIMENSIONS FLOW AREA RADIATIGE CONVECTION GAP HEAT TRANSFER FLOW DIVERSION METAL-WATER REACTION 4-HEAT TRANSFER
,,v v
CLADDING TEMPERATURE OXIDATION 1519 138
1A EF BREAK REFLOOD PWR REFLOOD AT FLOODi.1G RATES LESS THAN 1 IN /SEC. APPEARS TO BE WORST CONDITION BECAUSE OF APPENDIX K REQUIREMENTS FOR STEAM COOLING AND BLOCKAGE.
NRC PERFORMED LIMITED SENSITIVITY STUDY ON BLOCKAGE, STRAIN, AND INCIDENCE OF RUPTURE.
1519 139
SWELLING AND RUPTURE REFLOOD STUDY BLOCKAGE STRAIN t
T T
RUPTURED NO)E UNRUPTURED h0DE BLOCKAGE ELEY.
STRAIN TIME PCT ELEV.
STRAIN, TIME Case MODEL MODEL rup
, up
,INIT FRACTION IN.
SEC.
'F IN.
SEC.
jPCT
'F 1
WREM WREM' 29.C 876 1804.1
.522
[
80.24
.444 260.
Melt 33.51
.085 260.
Hel t 2
VENDOR WREM 29.8 6/6 1804.1
.362 80.24
.444 44.25 1824.3 33.51
.085 298.
2143 3
VENDOR 1.0 36.4 915 1932.0
.238 80.24 1.0 45.5
?104 F 13.34
.133 120.
1869 4
VENDOR WREM 45.
976 1780.0
.310 80.24
.394 44.25 1791.8 33.51
.088 298.
2040 5
VENDOR 1.0 45.
965 1763.6
.200 80.24 1.0 44.25 1773.4 33.34
.195 120.
1820 CD 9
a f GM f OYM l
3 I
g z
b b
I M
J h
c l
IE.
i I
I i
I t
1519 141