ML20138L518
| ML20138L518 | |
| Person / Time | |
|---|---|
| Site: | University of Michigan |
| Issue date: | 04/20/1978 |
| From: | MICHIGAN, UNIV. OF, ANN ARBOR, MI |
| To: | |
| Shared Package | |
| ML20138L505 | List:
|
| References | |
| FOIA-85-587 NUDOCS 8512190292 | |
| Download: ML20138L518 (39) | |
Text
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. t .a 2-gi REVISED 10/26/77 d
REVISED 4/20/78 License R-28 li 140 Gram Standard Elements Docket 50-2
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?- 70 Gram Control Elements Id j'
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SAFETY ANALYSIS d _;
I Ia 9;
p.3 6j UTILIZATION OF INTERMETALLIC URANIUM u
] ALUMINIDE (UAl 3
, UAl4 , UAl 2) AND
(,j - ' URANIUM OXIDE (U 3 98) CERMET gj FUEL CORES IN THE FORD NUCLEAR
'i REACTOR 4
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.j i
.i
-i , FORD NUCLEAR REACTOR 4
- t MICHIGAN MEMORIAL -PHOENIX PROJECT i ~
- l THE UNIVERSITY OF MICHIGAN I
Ann Arbor
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- } - June,1977 d'
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"a 4 Prepared For Il -
U. S. Nuclear Regulatory Commission.
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c1 *, 8512190292 851203 il PDR FOIA AFTERGO85-587 PDR .
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- l TABLE OF CONTENTS
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Page g
il TITLE PAGE i 4 i.
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E i TABLE OF CONTENTS ii
?
9 1. INTRODUCTION 1
-1 1
/ i i! ' :
- 2. OPERATING EXPERIENCE 1 j; 3. PHYSICAL CHARACTERISTICS 2
- 4. REACTOR PHYSICS 2 T
1 ! . 5. FUEL SWELLING 2 4
)!
o
- 6. . BLISTER FAILURE 20
$ ., 7. HEAT TRANSFER CHARACTERISTICS 20 1 '
- 8. FISSION DENSITY 26 1 . _
i APPENDIX A:. PEAK FUEL CLAD TEMPERATURE CALCULATIONS Ui FOR COMPARING MEAT AND CLAD DIFFERENTIAL
$ TEMPERATURES WITH THE CLAD -COOL. ANT O* INTERFACE DitttRENTIAL TEMPERATURE .
28 e..
n ; REFERENCES 36 a ,
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- 1. INTRODUCTION y -
$' intermetallic uranium aluminide (UAl , UAl , UAl ) nd ur nium xide i Rev 4/78 3 4 (U308) carmet bis differ from uranium - oluminum2alloy fuels in yj
- b. that the meat is formed by high pressure compaction of uranium-bearing powder and aluminum matrix powder into a solid. The compacted meat is clod in d! the some manner as a fuel meat casting of uranium - aluminum alloy. Powder fi metallurgy produces a more uniform dispersion of uranium throughout the aluminum
!i ,
'.j matrix and provides better controi of the fuel plate uranium content thon' can 2 be obtained with uranium - aluminum alloy.3
- This analysis describes irradiation tests conducted on aluminide and oxide 4 fuel plates between 1958 and 1976. Aluminide and oxide test plates were irradiated in the Materials Test Reactor (MTR), the Engineering Test Reactor M
(ETR), and the Advance Test Reactor (ATR) at the National Reactor Test Station Q (NRTS), Idaho Falls, Idaho, and the High Flux Isotope Reactor (HFIR) at j the Oak Ridge National Laboratory, Ook Ridge, Tennessee. Aluminide q test plates were irradiated in the German Karlsruhe FR2 reactor.
t The primary parameter examined in irradiation tests was fuel core swelling
(%AV/V) which could cause fuel plate failure.
i Post-irradiation tests were
]i conducted by heating of test plates until blisters appearsd indicating clad
.s failure .
l1 g
- 2. OPERATING EXPERIENCE r -
1: The MTR and the ETR have successfully used both alloy and aluminide fuels.
d .
Aluminide fuel has been used in the ATR since 1967. To date, the ATR has i ., consumed approxirnately 19,000 plates in 300,000 MWD of operation. Aluminide
?! ' - fuel has been used in the University of Missouri Research Reactor (MURR) 1; since 1970 and the Massachusetts Institute of Technology Reactor (MITR) since resumptierr of operations in 1975.
- y i 5 Since 1965, over 76,000 U,0 3 fuel plates have been operated to depletion at the HFIR with no majordifficulties. l U fuel elements developed suspected fision product leaks.On two occa
!- In one case, the
- apparent leak was so insignificant that the element was operated to depletion.
In the second case, the element was removed after 1500 MWD.3 i Destructive examinations of irrodfated test specimens and actual HFIR fuel plates have y p oduced no evidence of blisters, cladding separation, matrix cracking,
- ;, or any defects indicative of incipient failure.
?i j Aluminide and oxide fuels were developed and used because it was found that those materials permitted better process control a b superior in quality to those produced from ,exo,,pr.,in alloys.3 g,nd produced fuel el 1967 J
- at the ETR, o partial fuel element melting occurred due to a blocked flow 9-H ; channel. Although the fission product inventory in the fuel was substantial, the remarkable ability of aluminide fuel material to retain fission products 4 permitted easy cleanup of the reactor and retum to power within two days.5 i
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i :;- , e l 2-q Table 1 provides operating parameters for the primary users of aluminide and
}q <axide plate fuels and compares them with similar parameters for the Foni Nuclear 3 Reactor (FNR). ;Each of the reactors using plate-type fuel utilizes plates of Zi ,I ,
approximately fae some meat thickness, nominally 0.02 in, with clad thicknesses p$i that vary from npproximately 0.01 to 0.02 in. =With these dimensions being relatively con:, tant, it con be seen from Table i that the sigrMicant operating
- d. '
parameters of thermal power density, peak fission density, ona heat flux are ji
- more severe bp orders of magnitude for the reactors listed than the same parameters
'l 1 for the Ferd Nuclear Reactor. .
Tests referred to in succeeding sections were performed on fuel plates with meat lj and clad thicknesses nominally the some as those described above.
1-d 3. PlWSICAL CHARACTERISTICS n *
^ -
~r . .
?,
Aluminide and axide cores are not as strong as alloy cores. However, FNR elements
( are not operated near any high stress limit. A slight strength reduction is not
( t a serious armlytical censideration. The basic strength and integrity of the plate e.lodding remains unchanged.
- f. 1 Aluminide and axide cores are more molleable than alloy cores. In the manufacturing
' process, thickping F the fuel core at the e'nds during the pla't.-rollhe operation j' is eliminated. The result is more uniform clad thickness which adds to overoli strength and assures a uniform fision product barrier. --
The molting temperature of alloy cores is about 1560* F. N melting temperature
- of aluminide and auTde cores fs generally considered to be the molting point a
' of the 1100 Al metrix, about 1200EF. Except in cases of compiere flow blockage
" at high power or violent transients, none of these temperatures are expected
, to be approached. In any case, it is the disruption of the fuel clad which could a'
result in the release of fisicer products, which 'would occur ot obout the some s temperature for any type of fuel. Moreover, the type of event that could produce 5: such elevated temperatures is c.haracteriLad by essentially complete insulation
,i of the fuel by film bojljng wherein the vapor generated. completely covers the host transfer surface. The rate of temperetnre rise is of the order of thousands of d;~a per second. Melting of culloy core gould occur only a few milliseconds later than molting of an aluminide~or anide core, g -
1 u
j '. s 4. REACTOR PHYSICS - O ~
J No reactor physics changes will occur es a result of the conversion from alloy to aluminide or oxide fuel meat. The reactor core flux distributions and rod
.j wrrths will bkunaffected since all physical dimensions and characteristics will a, -
be unchanged.
n v -
'.ti 5. FUEL SWELLING q, --
y ,
l
}, . Alu einide and axide fuels Takhibir lower swelling rates than alloy fuels. In
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> _ TABLE 1 ,
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TRAINING, RESEARCH, AND TEST REACTOR OPERATING PARAMETERS
?
4 i Materials Engineering Advance
- High Flux High Flux i Testing Test Test Ford
[e isotope Brookhaven
$ Reacto r Reactor Reactor' Nuclear L Parameter Reactor Reactor Reactor I
l h - Year placed in service
_(MTR)I _ (ETR)I (ATR)I (HFIR)3 (HFBR)4 (FNR) p{ 1 1952 1956 1967 A-1 ,
1965 1965 Thermal pcwer (MW) 1958 U!
40 175 40 t.1 100 40 Thermal power density (MW/l) 2 0.75 1.2 4' 2.8 1.5 0.5 .025 fuel element meat volume (cc) 365 550 798 !ff,j 3475 870 U-235 per element (sm) 354 ti 200 400
! 975 2600 315 l-U-235 burnup (%) -
140 25 25 30.5 Peak fission density (fiss/cc) ,
34 35 1.8 X 1021 1.8 X 1021 j,9 X 1021 1021
! 2 1,24 X 5.44 x 1020 Rev 4/7El-Fuel element surface crea (ft ) 15 23 2 34 147 36 15 '
Heat flux (BTU /ft -br) ,. 3,5 X 105 5 X 105 4 X 105 2.5 X 106 3.8 X 105 Fuel surface temperature ( F) 3.68 x 10 Rev 4/78 '
i 239 329 356 300 304 159 Coolant flow rate (gpm) 24,000 t 44,000 16,000 17,000
~
..- 16,600 980 Fuel element materials:
Cladding i 1100 Al 1100 Al 6061 At 6061 At 6061 Al 1100 Al Ccre (wt% Uranium) 18% U-Al 22% U-Al 41% UAi Alloy and ' Alloy and x 41% U38 0 0% U3g O 14% U-Al UAl UAl Alloy X' X 9
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the former, swelling rate decreases with increasing uranium concentration due
];3 to the fact that the viod fraction of the fuel plates tends to increase with incre umnium content. For fuel plates with high void fractions, early core life swelling, is absorbed by the voids and the onset of swelling is delayed until later in core life.6 1
5.1 National Reactor Test Station (NRTS) Fuel Swellinh Tests
! Figure i presents swelling information from 1958 NRTS tests performed 4l an aluminide his.7 The percent volume change in the fuel meat is
%j > related to the peak burnup in fissions /cc of fuel meat for both alloy and aluminide fuels. As shown, initial void content in aluminide fuel
?- is an important pommeter in fuel swelling considerations. Expected d
f, bl core swelligwith no manufacturing voids is 6.38% AV/V per 1021 ry ranyce, 8 hpected miling curves for UAl x bl cores with d 4% and 7% voids are shown.
NRTS initially set a 7% AV/V swelling limir for the flat plate ETR geometry.0 I) The limiting swelling criterra was selected because some warping, attributed a to miling, was periodically observed in flat sample platelets irradiated to tb 7% AV/V swelling level. In the more' stable curved geometry of FNR bl, a 7% AV/V would not result in any failure mode. In
{
< fact, tk upper limiten %I plate swelling can be based upon- 1) Integrity of the fuel plate cladding and 2) Thermal-hydmulic considemtions d, such as flow blockage. in a practical sense, the upper lin.it can be
- the maximum experimentally or opemtionally verified swelling wherein 7
clad rupture does not occur. A high percentage volume change in the 4 -
W; fuel plate results in a very low percentage reduction in fuel channel area. For example, in the FNR geometry, a 20% change in tb fuel d
core thickness (.004 inches) results in a !).4% change in the'.117 inclr flow channel thickness and the flow an a. Such a change is not a significant
]d 1
operaticnal consideration.
As con be seen on Figure 1, the maximum FNR fission density results Rev 4/78 in a3.5% AV/V in alloy fuels and no swelling in aluminide fuels because dl -
of the void fraction.
Yi Th results of NRTS irmdiation tests conducted in 1967 on aluminide fuel plates in the MTR ore show"in' Table 2 and Figure 2. I For core fission densities up to 1.4 X 1 fissions /ce, swelling reached a maximum of approximately 7.5% AV/V.
y Additional results of NRTS irradiation tests conducted in 1967 on alum fuel plates in the ETR are shown in Table 3 and Figures 3A and 3B. I
.,9 Swelling in aluminide and oxide cores was consistently less than swelling q -
observed in agoy cores as can be seen in Figure 3A. Up to fission densities
% of 8.25 X 10 fissions /ce, the maximum alloy swelling observed was
$ 5.25% AV/V while tb maximum aluminide swelling was 4.5% AV/V.
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Fission Density (Fispec X 10-20)
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FIGURE 1 a;$-I 1RRADIATION SWELLING OF URANIUM - ALUMINUM FUEI[
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1 1 TABLE 2 A .
2 ' DATA ON SPECIMENS CONTAINING POWDERED UAl 3
-j i IRRADIATED IN L-51 POSITION OF MTR]
g : Somple Number e,i 2 Core .UAl3 in Hear Transferi U-235 Core
' . Composition Samplas Clodding ' Matrix Core Surface per cc of
- Number Tested Alloy Fission Densi
.. ; Alloy (wt%) (ft2) Core (g) 20 77,37c, m] . - (10 9 '
113 15 X8001 X8001 46.7 27.6 1.08 20
,,j 114 10 6061
- 3 6061 46.7 27.2 1.09 19 116 9 1100 1100 29.6 15.9 0.597 12
'1 118 6 XAP001 1100 47.0 27.6 1.20 19
'k 21, g.
Total = 40 r.
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- . O 2 4 i N! 6 8 10 12 14 16 8 rg Irrodlotion Exposu'e r Of Fuel Pfore Core-IO20 gg,,fge m.. .- . 3.+
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4' '.1 i 13 , FIGURE 2 jj j SWELLING PRODUCED IN POWDERED UAl 3 FUEL PLATE CORES i i f4 BY IRRADIATION IN L-51 POSITION OF THE MTRI 1 .! .
x.
e
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DATA ON SPECIMENS IRRADIATED IN J-8 POSITION OF ETR' P.';
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- j g3 Fissile U235 Oxide Sompte Number Core Calculated d Composition Somptes Clodding Matrix Fissile -
Compound per cc of Cootirig Plate Core
}p Number In Core Core Thickness Surface Tested Alloy _ Alloy _ Compound . (g) Fiss Densit (wt%) _ (mils) Temp f C) _(10 fiss/ce)y
. g., ATR1 4 X8001 X8001 U0 .9 .9 38 170 6.8 to 7.7
{' ATR 2 4 X8001 X8001 002 32.0 0.90 2
> 140 3.4 to 6.0 l'.i ATR 3 4 X8001 X8001 UA1
. n 3 35.4 0.78 2 170
':j ATR 4 4 6061 5.3 to 8.7 X8001 UA1 48.2 1.16 j 3 1 180 10.4 to 11.2 ATR5 4 6061 X8001 U0 9 38 .
.9 1 170 6.5 to 11.4 ATR 6 4 6061 j' ! X8001 UO2 41.2 1.24 1 i 180 12.3 to 12.5 ATR 7 4 6061 X8001 U038 .9 0.M 2
.{ ATR 8 4 180 7.3 to 7.8
- 6061 X8001 UO 32.9 2 0.89 2 150 5.7 to 6.1 ATR 9 4 6061 X8001' UAI 3 35.4 0.79 2 180 i : ETR 10 4 5.3 to 8.7 1100 22.wt% U-Al i
0.68 -
100 7.1 to 8.3 ETR 11 4 1100 1100 'UAl 30.9
. 3 0.65 -
100 8.2 4
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at u Al 3Powder Core =O 5! l 8 Espected swelling
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Swelling Of U.Al Alloy Cores
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Irrodiation Empesure Of Fuel Pfote Core-1020 ;gg,,f,e rec.-e-ape
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",, . FIGURE 3A
- 1.i
- _ SWELLING PRODUCED IN. POWDERED 3 UAl ALLOY FUEL PLATE CORES Dj
^ t }4 BY IRRADIATION OF SAMPLES WITHOUT A THERMAL BARRIER COATING
. . a, IN THE J-8 POSITION OF THE ETRI
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Urosivse Oside 10 l t Fueled Plates il i i
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g f rrediation Es peeste Of Fue6 Ptote Core -80" fiss /cc me-e-eye
- d}.-
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FIGURE 3B p , j SWELLING PRODUCED IN POWDERED UAls FUEL Pl. ATE CORES BY IR yl r OF ALUMINUM-CLAD PLATES COATED WITH A THERMAL BARRIER OF Al O e IN THE J-8 POSITION OF THE ETR CORE 23 [)1 S.
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g j i . e A .*- i Q{ The maximum aluminide swellig from Figure 3B was 2.5% AV/V fo fission densities up to 11 X 10 fissions /cc. v.2 .. g ,
@j -
Another set of 1967 NRTS irradiation test results for aluminide fue l ji in the ETR is shown in Table 4 and Figure 4.3 n e was approximately 5% AV/V up to fission densities of 1.6 X j 10
.j -.{ 1 fissions /ce. .
g'j Continuing studies at NRTS of reactor fuels and the effects produced ) by fissioning have led to a better understanding of the swelling mecha 6 Swelling is produced primarily by solid fission products. Figure 5 shows Q, calculated and experimental swelling data for several fuel systems. jj The strong tendency of fuel porosity to contain the initial swelling is y1 well illustrated. The inherent lower density of fuels made by powder metallurgy offers benefits in terms of fuel stability. kl , Experiment INC-16-1 was conducted at NRTS in 1971 on a single UAi i} f fuel element desi x
- x. r . were examined. pted as XA003F. Individual plates from the element q]
,'[ Nineteen plates from ATR elements similar to XA003F were measured for length and* width changes caused by fission product swelling. The
;p
' l- average
.13 % .16 length change was plus .22%; average width change was plus
; These values establish that swelling volurne changes occur almost entirely in the thickness of the fuel plate.
. I Measured fuel plate swelling, based upon thickness changes at three j)j;l: locations on three XA003F plates, is tabulated in Table 6. A plot of the L. swelling data, Figure 6, shows that, with the exception of sample 3-swelling does not exceed the predicted value of 6.38% AV/V per 10
$. - fissionvee. .
gl
'{ j - The INC-16-2 ~ experiment in 1974 consisted of nine UAlx ATR compo Gi fuel plate specimens being igiated inR the fluences of I.75 - 2.26 X 10 r/cm27,,,,
for 55,612 {7 MWD at ~l M in Table 7. Specimen data are provided q;
))- Measured core swelling values were considerably less than expected
- j M,
- j. -
by calculation. Figure 7 is a plot of fuel core swelling versus fission density for the experiment. g;} ,l. Visual inspection of the test plates showed surface discoloration but no Indicatfortof failure. Photomicrographs revealed the irradiated fuel .] w { to have a sound microstructure, w. ^j l r, .. 5.2 Oak Ridge National Laboratory (ORNL) Fuel Swelling Tests i; 0 i [.3 { Test irradiations were conducted at ORNL in HFIR to measure the performance and physical characteristics of aluminide and oxide fuel dispersions.2 i{ Fuel plate datcr are shown in Table 8. Swelling data are shown in Figure 8. yi ej T' he 3U 0 identified as powder blend, PB-01, is the oxide presently used or iflR. UAix blend,. PB-32, is a material similar to that used in -
- w{N the ATR.
' ;1 ., l . - . , - ~ . , .
;~.
. , .; .. _ . x..a p ..a: ,
-- - - -- .CYM$,7 7'$,d(($ ' ' ' '-
.,r.
---.~%.-. . . . ..a i- + :---.~s,7f.=
755 -"r.sv -9 '*.E'"
- ~ .$ .'S .' ., . 2- - ' - -
i . . 1 . .
'l. , . ;i '
i,,' : N , TABLE 4
-t
[j-. } - I 8 DATA ON SPECIMENS IRRADIATED IN G-12 PRES 5U $>; j q{ Sample Number Core Fissile U-23S y1 Composition Samples Cicdding Matrix Compound per cc of Core Ql n t Number Tested Alloy Fissile Alloy Compound in Core Core Fissig Density
. '4 (wt%) (g) _ (10 fiss/ce) a . ' i 11 606i X8001 UAl '48 3} 3 1.17 7.1 to 19 i; 2 6 6061 X8001 UAl 52 !i 1 '
3 1.31 6.4 to 14 y' ; i 4 j 4 l 'l 1 - 3 1 - a Il .
! i 4 1
, IO ' '
AQ $ - Espected swelling -
'4 ' s \
$ ! E j S I
.-.- L e
I
- - a '
f
; 4 -
j s. g o. -- . . p 2 ' e2 .
- 2
.5 4 y
Observed swelling -
'{
- l' 4
1 / j
/ . 7
/ "
i
- ,.i 0
/s- .
?
0 2 4 ;
' 6 8 to 12 i 14 16 88 trrodiotion Espostare of Fuel Pfofe Core-10 fiss/cc M '*****#
d FIGURE 4
] _ SWELLING PRODUCED IN POWDERED UAl, FUEL PLATE CORES BY IRRADIATION IN G-12 PRESSURlZED WATER LOOP I OF ETR l
{ l ql <; e -- . ~ -
.: m .a.m u w .x a a=:...z =.: ....:.._. x . . z.: .:.u.m.v.a.:..8:.a . a w w .u..u w ._....'
, . =..._. _._ _ ... ,
.. . . . . . _ . . n _. ..
- . la '
t, li i 19
. . Lia.
14 - - g Swelling.
.I . .
l Y,
.. .:.<;. ; per'atoa %:-
U:.. '.. Test' i ll Fuel - burnup; Density
~ ~
Temperature - 4 12 c' Dense UO.2 . . .
'. . l . 7 ', ~ ~ . 10,5 - -
<70000 t
~ .
! U ' ' .... '. 3,0 . 18,9'-
,. <300cc -
s . 10 - 22 w/o U Al Alle'y 0,09 0,7 Calculated 1
<20000 - Susiling b.
t - . ni
. P.
. . t-
. li .
8
* 'P'j 8.,
1,,
"o . . . . -
Dense UOg <- - Uranium Metal X h>' . j.j g - I
. m"1.
- - s z-
/ i
..,' / ..;
/,
4 - i
- /-
- /' - -
1 g., - ,
-" t.
9
/
1
- y 4,
#f .
<.: ' U-AlAlloy
/
/ / -
/
/. / '
- t
/
/ /
/ ;,
l .
' porous UOg 4
.v s'
1 / l
' l , Powder met. U-Al illoy
} .
, i i l
- l. .
O 2
- 4 6 0 -
10' 12 14 16 18 l ..- _ _ . _ _ . - l IN C-B-ll206
- -- F. ISS/CCx.l.O. -20' . .
FIGURE 5 , FUEL SWELI.ING INDUCED BY SOLID FISSION PRODUCTS 5
g..pyr.~ _: r.E.U.Tdll?.L.
._ _ :.17 72*~ % ? '.L . ^ *~ ~ ~ "
- --- ---. - . 3
- t ,
. L,; 4.c .i..L.
Q; 1 m. 2 TABLE 6 I
- l. :i A
- h. ,
SWELLING DATA ON UAl x FUEL ELEMENT XA003F IRRADIATED IN ATR 0 ] Distance u: From Top Fission . Core il ', Fuel Element Coupon of Plate Density $ welling 9; __ Plate Number (in) 20 (10 fiss/cc) (% AV/V) d 1 1 3 1 2.5 1.6 i 1 2 7.I j 1 3 14 25 9.I 2.0 3.0
'; 3 3 1 3.0 1.9 l } 3 3
2 14 8.4 5.5
,.i. 3 25 11.0 9.0 j' 5 1 3 3.4 2.2 5 2 14 9.8 3.0 g-
.t .
i $ 23 15
, g: .
f
'y-4 :-
,m A
6 . 18-ww:.' 5 to d-' 8 Ei H :; - F bi e g.. d--f
- 4 u yy M 4
-i 15. -'
@' ~ -
1
'C
. a, F~
b, u, g 5 --.qZ . d .E' f u -
,- : .c
- r :.
p u.a
- w.
o
-- ^
-',u -.-
p 6 e . lj g . . .-) j 0 5 10 15 20 L.* Irradiation Exposure of Fuel Plate Coupon (10 20 77,,j,c) d, d.
- . FIGURE 6 i 16 SWELLING PRODUCED'IN UAlxFUEL ELEMENT XA003F IRRADIATED.IN' ATR O ' ,
t
- c * * * ~
,e s
l Ad
. o., .
i s 7, .
'TABl.E 7
.'j; ,
DATA .ON UAlXATR FUEL. PLATE SPECIMENS IRRADIATED IN ETR17 O U-235 Uranium Plate 9; ; Perccof Weight in Surface Percent Fission i, - Plate Core Core Enrichment Core Bilster Temgerature Burnup Density Swelling M q .; Number (gm) _ (gm) (%) ( C) Temperature (%) (1020 fiss/cc) (% AV/V) (~C) 4; 169-4 1.51 .7468 91.86 '100 - 260 82.6 26.3 2.0 565 i 169-5 1.51 .7286 91.86 100 - 260 90.9 28.8 4.7 >565 -
.gY 169-11 1.21 .6715 91.86 90 - 340 90.5 23.1 4.7 538 !
-j - 169-12 1.22 .6747 91.86 90 - 340 95.2 24.3 5.9 l
565 1i 169-19 .953 .5424 91.86 90 - 270 98.3 19.7 4.7 565
}! 169-36 1.21 .6554 93.10 90 - 340 98.3 25.1 I' '
6.4 >565
;- 168-37 1.21 .6772 93.10 90 - 340 99.9 25.5 6.0 538 '
j3 169-38 1.23 .6967 93.10 90 - 340 97.0 25.0 7.4 538 i
.[f 169-39 1.22 .'6651 93.10 90 - 340 93.3 23.9 5.7 565 4
- q ' l i! $
s [ t it
. I p
1 - ( 1 . i i ..
- i . .
, ,! I-
..g._..g..;,.......,..,..,m.,..
..,.7z...,.,..-.___.y.._
g.,-...._. . . . . .,
.._..,m.......
s,a i . .:r"i :
', 42 $4., :
..a 10 l
L !h O l l l M l Gj
- 1 i Il 1 L 1 .
J l' [
! l i i
\
m i I L
- a > " ,
.:j N g l.
l f
' : 1 sM.4 D i- _il: 1.'
l 1 8 n . l f j -l 'l i
- 4. A! 7
; $j !d il '
l l!!- l:3 . , ' I,. ' 3
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s3 l , l ' l.l '
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- a. j > d
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"b gy -
W l i b 2 q *
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v, .' l-i
@ ,7 W. u d Ei' i '
- ( h-t s 4.41 e 5 : : '
o [ j. Y
.in l,g GJ W E c!g __
e a 4 i a4, -+
&u;i u. *:. l' tt F '
..t-J ] i u;h r .c :;i.
; r i
;. .g l
- ';.M
}
e 3
- I"{;
I I gur. ' ' i l l ,'M.t -e m i- ; i ,V' ' c ,J
- t. .
.8 i J l 4
.ip u 2 ,
,, ^
- :i ,,
! If 1 L_lh, !
- i;l- - -a i A l l,
b J ia . 1 , c l ,a'i % 5 10 , 15 20 D 33 3
, 6 Irradlation Exposure of Fuel Plate Core (1020 fispec) o
- }lcj; FIGURE 7 kI a
,1: II _ swell. LNG PRODUCED x- IN UAl ATR FUEL PLATE SPEClMENS lRRADIATED IN ET
.\ ,.
f ' '7Ii . q. . - S d
- j .
~~ ..,- - _?7. :.qis*mz n .m]& * '" 7' *? T."T"' ~ T'TW ' ~ ..,,.7 . . .
-- 'c 'r' r' '----~.:
' ' ~. - , :_ X- M=-- L'M ' 2 ==--
a
,.:_..___ g . 3.{, f E psJ~ ../ j.
,;7;. 9gg gj.. n,;g, g, . , .u.. ,
t . . 1
- j . .
'A. 3 , TABLE 8 ? 9 k-
- PERTINENT CORE ATTRIBUTES OF FUEL PLATES IRRADIATED IN HFIR IN PM CAPSULE 12 h:3 !
l.) i P! ate Dispersoid n i , Uranium Loadi.ng fj t n l Void Fissile (.f j Concentration Content of f, - Powder Density i Blend .g.YP* Dispersion Total (g U-235/cm 3 3 (wt%) (vol%) (vol%) (g) of core) g .- 11-3 PB-33 Arc-cost UAix 53.2 31.7 3.1 1.088 0.00 3, 21-4 PB-33 Arc-cost UAix 54.0 31.9 4.5 1.089 0.00 M' 31-1 PB-33 Arc-cost UAix 54.8 32.2 5.8 T.088 0.00 12-3 PB-01 High-fired U 0 47.1 38 21.7 4.1 1.187 f.41
! I' 22-4 PS-04 Bumed U O 49.8 38 23.8 8.2 1.187 1.43-4 32-4 PS-32 Arc-cost UAl 51.4
' X 30.3 1.3 1.085 1.31 13-4 PB-01 High-fired U 0
' 38 . .5 3.0 0.9.53 1.14 23-1 PB-04 Sumed U3g O 42.0 19.1 6.4 0.954 1.15 i [ 33-4 PB-35 Arcw UAlX 53.7 31.6 4.6 T.087 0.00 4, - 14-3 PB-32. Arc ~ UAIX 52.9' 30.6 4.2 1.086 1.32 2 24-3 PB-32 n; Arc-cost UAl 52.2 30.3 3.3 X 1.085 1.31 L' .i 34 PB-36 Arc w e UAI 1
x 53.2 30.9 4.3 1.089 1.32 ji 15-4 PB-32 .. ! ; Arc-cast UAl 62.8 38.6 6.6 1.344 x 1.67
)} 25-4 PB-32 Arc-cost UAl X
64.1 37.3 8.4 1.345 1.69
~ ,'
35-4 PB-36 Arc-cost UAl 63.0 38.7 X ' 7.3 1.342 1.66
+
T6-2 PS-II High-fired Uf08 '* - y 26-3 PB-34 High-fired U 0 .9 38 . 0. m OM L [.i g' $ 1 .. -= j j . il , }, < 1 f N i
- I ~. : .- a- -
x
. . 1,::My*.QQS.bp$$?::.;5&.
.u.._... --. m_u .N.. g.jl7 yJ.1%. ;
,,7.
- . ;.n.m.y:;.7.m.3g r:.
. m._
.^
17_ bf q.I , .
~ 'f . -
(6
. CRNt.-DWG 72-12367 7.} g ,
} , A. ARC-CAST UAI, DISPERSIONS j ifj o HIGH-FIRED U O 3g DISPERSIONS
'l y
- BURNED U O3 g D!3FERSIONS ,
. .] a - *
- ri 9 .
12
'/
'j
. i cat'.CULATED SWELLING RATE
'd i W . -
FOR 400 % DENSE FUELS 2.i E . 432-4 - ci $ E, u 8 r 24-3 -. g4_3 4 W ..
.l 2 - '
3 - 34-2A *
-) '
a C3
. - (5-4, W 1
~ 35-4 '~ r'
$4 (1-3, t -q,33-4 21 -4) 13 - 4, 25-44 ."
.j 5 Y' gj .pt2-3 23 -4 22-4
. u .
i.-> s
-i ' & 16 - 2 -
26-3
- ) -; , . .
'i1s - , a ~- q -
.~ ,-
. i
'. /
.4 .
i th ; o. 4 8 12 16 20
. i 3)-
(x1020) }.
,, SURNUP(fissions /cm ;
r, - l-l. 3 FIGURE 8 q;. kj
SUMMARY
OF FUEL-CORE SWELLING DATA FOR MINIATURE FUEL PLATES l. IRRADIATED IN THE ORNL PM-1 EXPERIMENT. DATA POINTS @1 ARE 10ENTIFIED BY THE FUEL PLATE IDENTIFICATION NUMBER 2 t.d LJ e l' I sv
- g. ,
\l . I ___ _l____ _ _ i-- -
$$$bkYM7.E.us:'nWST"'WJZ E EE' ~~ ~~ '
d . .
} , , .
Li Q . 1 All 17 test plates were inspected visually in the HFIR pool appro
- j. 24 hours after shutdown of the experiment. The plates were free of warpage and distortion and generally appeared to be in excellent condition.
d9 The test plates were allowed to cool for approximately three weeks in
!.5 j $ .: the pool and then transferred to the High Radiation Level Examination k Laboratory (HRLEl.) hot cells for detailed examinatJon and evaluation. % Examination consisted primarily of visual inspection of the plate surfcces b
for indications of structural damage or defects, determination of the plate weights and densities both before and after chemical removal of corrosion product films, analytical determination of the speciment burnu and metallographic examination. L 1.
# 4 Y j- in general, the examinations showed all 17 test plates to be in excellent condition.
f The density and weight-loss data indicated that the plates
% 1- had increased less than 0.002 inches in thickness, including the oxide A film formed on the plate' surfaces during irradiation, and that corrosion f , ! had reduced the thickness of the cladding of these plates by less than ; 1 0.0002 inches. Extensive visual and metallographic examination of
[. sections from each plate revealed no indications of actual or incipient- [; structural failure of any of the test plates. Consequently, it was concluded
} that all 17 plates p.dm.. d quite satisfactorily under the imposed condit i of burnup Igvels of I.8 - 2.2 X 1021 fissions /cm3 at irradiation temperatures I of 60 to 98 C. Slight bh significant differences in the swelling of
$i the various fuel dispersions were observed. As shown in Figure 8,' p i containing aluminide dispersions consistently swelled more than plates 5
. . ,! containing the 3 levels. U 80 dispersions when irradiated to comparable burnup 9i it,* .
5.3 German Fuel Swelling Tests _ i In 1976, UAl 3 and UAl I"'I P in the Karlsruhe FR2 2 reactor.Iglates : listed in Table 9 were irmdiat Maximum burnops attain ll.{ 4; to 72% which correspond to fiss. ion densities of 1.8 X 10 I were 53 ) fissionpcc. to 2.7 X 1021 D g,3 j Fission product swelling occurred at a lower rate bn the expected 6.38% AV/V per 1021 fissionyco. Measured values in Figure 9 show f ?, Mt, up to approximately 40% burnup which corresponds to a fission density of approximately 1.5 X I fissions /ce, the swelling rate is approximately 3% AV/V per 1 I f 2 fissions /ce. Above fission densities j : of 1.5 X 10 ' fissions /ce, the rate of volume increase was considera greater. The swelling rate in UAl 0 ' l 2 P ates with identical uranium content h .j was somewhat higher bn that of UAl 3 ; the difference is not much greater
- d. Mn the largest sectrer in the measured values.
i i i ,I ,* . !s? Nl - Ma:w~~~---"""~~~ f_ m
a ' #1r:x:1.h . : - -
"1*~:' :.ar :D?. .. . ~.T ' =
. . . . 2... a.. -. .=... . . .. .' . '. ' . E. '..s j
.h . d
'1 !
TABLE 9 N: h.. , O URANIUMALUMINIDE-CONTENT AND IRRADIATION CONDITIONS 3 FOR THE FUEL PLATES [q t UAix Content Mean Surface Mean i 9 - of the' Dispersion Temperature U-Burnup fl Test Rig No. No. of plates b (wt%) fC) (%) Irrodioted
'?-
jh .jj 1 2 50 UAl3 70 16 4 50 UAl3 70 47 9 3 4 50 UAl 3 g, rq 4 50 70 70 34 4 5 i UAl3 26 1 45.5 UAl 70 l 5 45.5 UAl 70 26 3 )Y I 6 '50 UAl 3 70 21 M 4 2
~~
.{ t 70 44 j 45.5 UAl UAl2 -
2
, 7 50 70 6.5 i - 3 ,
3 45.5 UAl 70 6.5 9j 8 50 UAlh 28 1 c 9, '50 UAl 3 120/ 135/ 150 4/4/4 h)-l s 45.5 UAl 2 150/16F180 53 2/7/2 T50/165/I80 53 . 2/7/2 10 50 UAl 3
'( }j 45.5 UAl 2 150/165/180 72 2/2/2 150/165/180 72 2/7/2 TT '50 UAl 3 d.) j 150/16V180 60 1/1/2 45.5 UAl 2 150/ 165/180 60r
}.4 3 54.5 UAl2 T/T/2
- s 150/165 60 2/2
- I
- :
ai 3 ;
'1 ; '
3^ 3 m. m s .. + sus.=r-a & '""
+
L s msm, a s > ,j .
'll ; {a- . ,,, .
$g ' l
' a EZI- . , . . ,' - !
i - ETA J ! Is-1 F - L- I.. j , T
. . , as ..
pt ^ j
> . . .. s,- , - . FIGURE 9 d -1 )
Jl SWELLING OF THE FUEL PLATES (RELATIVE TO THE FUEL \ NI g; DISPERSION VOLUME) UNDER 1RRADIATION AS A FUNCTION ) OF BURNUP AT VARIOUS IRRADIATION TEMPERATURES 15 - I l; -t l %, llI gj ,g/ mh8." * - - ^ - - 6.'-*- *d _.
~z - .y 2 . _ . ,
--a _ . rx.J.u
.~
. . _. ;,,._ w. . u. : ._ , ._ t.. . = w ' , s, .
3' 0 6 BLISTER FAILURE N 11 ~ - 6.1 y:; : NRTS Blister Failure Tests D ,i j' irradicted cluminide and oxide test specimens et NRTS were heated
- .j]; in oir for 30-minute periods starting or o temperature of 260 C.I After each heating period, specimen pictes were inspected for
)Yi j ' and, if not blistered, heated to successively higher temperatures. The results are shown in Figure 10. 1 y: Figure 11 presents blister anneal data for aluminide fuel. " For a giv i [ .; . peak fission demity, the curve relates the minimum temperature at ! s
- ,' which fuel plate blistering may be expected.
i 1 7; ' Pbst-irradiation blister failure test results for experiment INC-16-1 y are tabulated in Table 12 and plotted in Figure 12.16 Annealing tem was raised in 50 F increments during the test. in general, blister d 3 tempemtures are within 50 F of the failure - no-failure curve established >j for UAixfuel plates. Matrix-cracking was the predominant blister M i mechanism, though some microcmcks were observed. in the fuel. 4 :,
. All fuel plate specimens in the INC-16-2 experiment were blister tested following irradiation. The results are tabulated in Table 7 and plotted d 5, in Figure 13. A previously established UAix blister curve is also shown in Figure 13. The INC-16-2 specimens blistered well above the blister
,, constant beyond a fission density 21f;,,;,n,7ce, of 1.5 X 10 tempe a,
, 6.2 German Fuel. Plate Temperature Failure Tests
-t i 1'
jl The test plates listed in Table 9 were annealed for one hour from 200 - jj 500 Cin steps of 100 C following irmdiation. Volume changes were measured after each step. j: - Up to 300* C no volume changes were evident. At400 C, the first 'i
. volume changes occurred. These took place only in the sompte plate 1( ,,
which were irradiate.d at the lowest mean surface temperature of 70 C. r., 21 at 500* C. The sample plates mier showedirradiated pronounced swelling at 70 Plates
}' C blistered at 500 C annealing temperature. In all other cases, no blisters were visible.
Ai ?
- 7. HEAT TRANSFER CHARACTERISTICS x; -
3, Rev 4/78 In Appendix A, the peak fuelglare temperature during reactor operation is calculated to be 156 F or 69 9I "$. the same value because the temperature rises in the meat an than i F. / .a. , 2 The thermal resistance, K, for the powder metallurgy fuel meat is appro v,t - s.. Ih* 1 { . __ _ .
- L_.-.. .-- ---- -- - :-
.. ..:. . . .... .w- -.w-
....... x.-.=:.. ... :. u. a.....x..
.,..:.i..._.
f 1.j i . 700 ' 4
.' ..1 .
3 i i-
..M .
.; j
. 600 _C -
- - - -- Powdered UA1 3 i
) .
- - / xy ,
, E* 500. J
'=
/ s
/
, W 'I 9</ ,/ 7-i e
(! b //,/ !
-i
_.= a 400, ( i w\ A - w N
*A(
Ah A A K
% y U038
. 3 e
300
?
_a- a i i 2 -
- 7 2 002
[ E-g N .. e- -
.i
. 200 a.
,1 Fuel Materials, 1 i
-g Powdered UAl3 =
- o
' 100 ..j [ U-Al Alloy = o l I
ai y . U038 -
= A 002 = 4 d? O l g , , , ,
c' O 2 4 6- 8 10 12 14 16 18 20 22 a. 2:e - n s, FueT Plate Core Irradiation Exposure - fiss/cc x 1020 I! ;
- PPCo-8-8382
, 1 M.!
[j , FIGURE 10 Qj RESULTS OF POST-IRRADIATION ANNEAL OF SAMPLE FUEL PLATES j' 1RRADIATED IN L-51 POSITION OF THE MTR COREl I I' I 'i {' ! L -_ ~
~ "Tv:% " ; ;q m>ML.'*
- a. .-_ ._____.--.mm"=:==>- -""C-""'- ' ' ' ' ' "
~ ~
"'" ^^ "^^
~~., ,
,-i. F.]' . . ki 1; 900; i i
;u . .
- .2 - .
* ~ '
O ATR HOT. SPOT CHANNEL O 800- - .
.' O ATH MAXIMUM NOMINAL d -
, q v .
O ETR HOT SPOT CHANNEL E i . g d., - ' . O. ETR MAXIMUM NOMINAL- - Hj.: t, 700-
- A MTR HOT SPOT HOT CHANNEL
. ;a ~
A MTR MAXIMUM NOMINAL 1 0 ,kf - f 4 , o g g y 4 .
. s . - ,
o -
. . 6
%.ss s .
. 'e. 2' 600- i s4's s - ' f.j { ss MTR & ETR Alloy (18 to 22,w/o) , , (
., E 500~-- $ ,
s, ' ~
' ' ~
. y, lo- ! TR Regular ($0 W/o)
- L t
^ f
<j fl ATR "7F" (60*w/o) O \
.o. Only One Point Determined
- ETR UA1 ' '
q
- a. X (31 w/o) -
>t i; f 0 0l-- .
~
m i ETR Foi1.& Alloy (22 w/o kith Doron foil) m A u 3 i l . o N0' FAILURE EXPECTED - g
' 9' 300!--
i i .i -
. l, T .- x
+
l - 1 d - w
;! 200- .
The 'kaximum Nominal Burnup Defined As The . g f O O I Peak Nominal Burnup And Temperature In The
;j -g '
A i
- a. 4 .. Mottest Vortical Pinne In The Hottest Element 10'O- ' 'E ' ' ' '
4.
! O 2 ,6 8 10 - 12 14 [
16
~
18 20 22 24 d Fission Density (fissions /cc x 10'20) l
.i FIGURE 11 P t .
J I. i . -, . . _ BLISTER ANNEAL DATA FOR ALUMINIDE x (UAi ) FUEL" , a; .~-.r.-.-.~..-r. ' " ' *
.- .~
~ " ' ~
. t. .-~~~.~-m--w----+-------
-*--e-~-~~~~-----~~'~-
. . ,_ .; ,.,. y --.. . y ,, % . e,
' --- g s,'i,m ,=:. v + -* -r r , pe % - V
- V","' C*.**T"7 ' ~~ ' '
'a*-
_________J__2 . 4 ' , .
.. . _ -.sx ==i,>:..nz 2 .
.. i .
__ h r4.i m . t .5 y 5 y. m 3 ._" "
. . (_;_. _: 2 ; 5 _ ? r-. -4 .
1 .
.j .
q . TABLE 12
- .A.
BLISTER DATA FOR x UAi FUEL ELEMENT XA003F
',: i a
4
;_ ). Estimated Peak 9 Exposure in Coupon Blister ] Plate Blistered Sample Number 20 Temgerature
__ (10 pg,,7ee) Y; _ { p) 8 T 16
> 1000
- j !8 2 13
> 1000
; . 8 3 9
>1000 j;]4 15 1 18 850 15 2 J[. . 15 900 15 3 l.] 10 37 1000 3 ~
I 17
?i 17 950 2 4 1000 -i 17 3 9 r 950 18 I ~
14 p, 18 8b 2 12 900
. 18 3 8 950
[ 19 1 15 [ 19 2 12 900 900 19 3 (it . 8.6 930 3';
,.4 M
d
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M. . i. . , l l
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i,
. 34 600 ,\
44
" N '
l .:N .
~ .
Ni
=a - .
,13 - 1000 .
t T O 3 O g O
~.. id, G o" 500 - 950 F O C3 O O 9 '
.!j.$ tj
- 900 F G O tl O
$ 850 F ' '
OO 9I
- s o
O. O 9
% - 800 F w O O l] g 400 -
3 } .
] - i a. , ' ' j, "
O 0.gs 235 * , i 0 gfee, 4 Blistered (represents temperature d 235 at which blisters first appear) - 3 i i3 0 1.2 U/ce 300 - L .-E '
- -m * -
0 Unblistered (represents last 0 1.5 g'235 U/cc
- ;.; temperature tested before g l- Q blistering) 200 ' ' ' ' ~ ~ k
, 0 5x1020 10x1020 15x1020 20x1020 l .. i 25x1020 ' ! l' Fissions /cc ; j- jj ti 1: _ t i i ; '
; a FIGURE 12 '
!7;: o ,! 4 - _ BLISTER TEMPERATURES FOR UAlx FUEL ELEMENT XA003F PLATES SUPERIMPOSED ' d u- ON FAILURE - NO-FAILURE CURVE ESTABLISHED FOR UAlx ,, d
+ .
SAMPLE FUEL PLATES ' p ." k z
- " ~ ~ ' * - ~ ' ^ ~ ~ ~
, w ~ p e,4 .** y 21
*4
**7.C,,'
'I [ I' j _~[. . ' , _d ,,
- -- I
,.;=;1.z u gnr:; m 2.3 C B ':C M a.u::sE1..-.'C Z:: ? -
~~~ ~~
li
.4, u
ii - 700 L L . L .
$ . 1 1 9.a. - *
., 1 . I
.. . { . -
p
;~
O 00 0 O O-y O O W96 0 - 500 ~'
- O Go -
.o .
- s. - .
.a a
- 400
.:i e **
.i -- -
- w -
t. z. g Blistered 3 300 .,, . .
-l O Unblistered. (repres"a'nts
] ; .
. hst temperrtwe tested.
a i . hefore blistering) +
~'
i 200 - I ! 't 1 . - I O 0. 4 "4
'.} 1 oxio21 1,gg7oz i
.i 2.ox.1021 2.5x1:21 3,e=yg22 Fissiondensity,fiss/cd
- a ,
2 g g FIGURE 13 l Jj 1
- POST-IRRADIATION BLISTER FAILURE PLOT l f3j FOR UAlx_ ATR FUEL PLATE SPECIMENS ~
] .. _lRRADIATED IN ATR'7 i I s I ri 1 i \ ' l, ;- - - . ! \' . . w s w - _. -. - -. - , - . - - - J
- &.h..:,m.u ..
.-.x. - -. , ~ -
3 an=xw.1==a:.ma::w. an TUa ~- ' i - -
- j. ! i , . '
-l.
fj , 90% of alloy. thermal resistance.10 The conversion to aluminide or oxide meat ['; will result in approximately a 10% rise in the meat differential temperature. J
However, since the total meat differential temperature is less than 1 F, as Rev 4/78 compared to a 36 F clad - coolant interface differential temperature, the j ; effect is inconsequential.
Kj Oxide coatings of up to two mils buildgp on fuel clad during fuel use. ' q! A temperature rise of approximately 16 F across the oxide coating could result g; Rev 4/78 in increased maximum fuel plate temperatures of approximately 172 F.4 h( Rev 4/78 The FNR peak fuel clad temperatures of T56 F and 172 F with oxide coating
'l; l are well below failure temperatures for aluminide and oxide fuels for all fuel
. fission densities for which data was taken.
; 8. _ FISSION DENSITY H
j" ~ Fission density limits are based upon acceptable fuel swelling limits and failure o - due to fuel blistering. Table 13 provides a summary of fuel plate swelling and blister failure dato i from Figures 2 - 13. The swelling dato provide the maximum fuel plate core j]j . swelling observed up to the maximum fission density listed for the test dato
- i ~ reviewed. Similarly, the blister dato provide minimurn blister failure. temperatures 2
observed up to the mcximum fission densities listed.
- j. Rev 4/78 For pack FNR clad operating temperatures of 156 - 172 F, no fuel blister failures were experimentally observed.
The maximum fuel plate swelling observed,12.7 % AV/V, reduces the FNR 2
! element coolant Flove charmel by 2.2%, from 0.117 inches to .114 fnehes,
; Rev 4/78 a reduction that is operationally acceptable.
s
.,i > '2 FNR fission density limits should be based upon the dato in this analysis and the i peak values routinely achieved by operating reactors. The following fission density 1 limits in fissions /cc are requested for the FNR.
r i
- . i M,.
a , 3-. Feel Table 13' Operating '
;! Proposed Material Test Dater Reactors
- 1 -
FNR Limit T.]>
- Rev 4/78 Alloy (U-AI) 2.0 X 10 21 2.0 X 102T (GETR) 1.8 X 10 21 f' Rev 4/78- Aluminide 2.88 X 10 21 (UAl , UAl ' 1.8 X 1021 (ATR) 1.8 X 1021 4 3 2}
h( Rev 4/78 Oxide 2.0 X 10 21 1.9 X 1021 (HFIR)1.8 X 10 fl ! UO) 3 g N t
- j !
. .i )
! l ,i l s.
- w n -e m-
- e.. ..apeo G:ks
; . --, ..,,,, m 92g ;7 ;;;,+<-,, ;55.g3h::;mi&
2.'. * ~ ~ ~ ' ' ' 2. : .'. I .'.:. . c.d. . .1. .
. ~l . . ;) - ]! . .
M , TABLE 13 il i
SUMMARY
OF FUEL PLATE SWELLING AND
~!~
BLISTER FAILURE DATA
'j Fuel Plate Swelling Data \
a i
. l_ ~i Maximum Core Maximum Fission % Swelling Observed d, A
- .1 Fuel Material Figure (% AV/V) Density (10 2 rg,ttained
,fec)
- (
ij U-Al Alloy 1 12.7 20.0
.d WAl Alloy 3A 5.3 1 8.5 w < .4
- I UAl 2 a x 7.5 14.0 UAl 3A 4.5 x 8.'O
- 4 UAl
- g. x. 38 2.5 11.0 1 -
UAl 31 x 4 5.0 16.5
'.,i , t. UAl x 6 9.0 T2.8 UAl
- x. 7 6.4 28.8
.' UAl 8 9.0 x 22.0 f- 4 -
UAl, 9 5 17.5 90 C)
-3 j U0 38 8 4.0 20.0 M.
I.3l! 3 Fuel Plate Blister Failure T....+;.no Data 3.
~
. ;. ; Minimum Blister -
);
^
Maximunr Fission x Fuel Material Figure .Follurg Temp) ture .Densg Attained 3 (~ C/ F (10 fiss/ce) r1 j U-Al Alloy 10 600/ 1112 11 '3 UAl 10 500/ 932 20 UA!* 12 j UAl* 13 450/ 842 550/ 1022 29 18 di U0 38 10 W 716 10.5 y); - u 2 10 3 q/572 i0.5 r3 1
) ,
i y .. . ,.
.- O
_- .__' m
. :. _ .. z f
._.3..._.._.
~_x .z.:.
__m...;..._.. :. .c su _ .. , ,
..Qy;,,,.,o
- 9. .. _-n ..a-
..qa@i .;y .
, q,,;w 1
g , . Li . a j
, a
, -i
) APPENDIX A h)i a PEAK FUEL CLAD TEMPERATURE CALCULATIONS FOR COMPARING MEAT AND CLAD DIFFERENTIAL TEMPERATURES
}j WITH THE CLAD - COOLANT INTERFACE DIFFERENTIAL TEMPERATURE
> <I! ~ 1 A.1 FUEL PLATE PARAMETERS A.1.1 Surface Area U
J] A 25-element core consists of 21 standard,18 plate elements B, and four control rod, 9 plate elements for a total of 414 fuel plates.
*4 y , A,= 414 2(w X l) (A1) o -
fl where w = Plate width, .21 ft
;j " l = Plate length, 2 ft e
1 , 2
- A3 = 345 ft 2
4 .t j , A.1.2 Fuel Meat Volume i
- j. Vm= 414 mt XwXl (A2)
.3 i where t- - Meat thicknesr, .00167 ft-j m a .
3
]; V,= . 290 ft (A3) 3 ha = 50! In ?) '
- 7) ' = 8212 cc/ core .
a i L 1., . Rev IV77 ,= 358 cc/18. plate standard element r: j Rev 1777 = 179 cc/9 plate special element I.i e Et A.1.3 Core Volume N)
- t. A 25-element core is configured in a 5 X 5 array of elements.
y Each element measures 3 in X 3 in X 24 in. ij
- y. 3 VC= 3.125 ft (A4) 3' 3
=Nn e-r el W9-- ,,. h e- 14 4Ta *_ _+._w- 1 -
u .
- .: = i & .= 2 t+ y : h & s r u r u ..= .x 2 r .. r S % d.rll,' Rv2 . . &-,. ax.
,2.
25.h G 'l +'-,
~'
3 i
; = 88,490 cc 9 ,a = 88.491 ?
- 4 j; ,a A.1. 4 Core Flow Area .
N! j A=N C wC Xt C (A5) C dl where N = 414 coolant channels $' w"= .218 ft channel width J.' tC= .0098 ft channel thickness (' s: AC= 414 .218 X .0098 d' - - 1 2
= . 884 ft . A.2 REACTOR THERMAL POWER DENSITY 1
I TPD = 2 MW/Ve (A6) q
}' = .0226 MW/I ,!' . A.3 REACTOR FISSION DENSITY l
j' Each standard fuel element contains T40 gm U-235 and is 17% burned up l at end of life. Standard fuel element bumop in gm is i BU= .17 X 140 t, (A7) i .
= 23.8 gm
- 3. I The averoga fission density in a standard element is
-g A BU Ao a t
- FD =-X X j ; ave vm N 235 af+c (A8) '
t where V = 354 cc/st ndard element A" = 6.02 X 1 3 atoms U-235/gm-mole
, d = 235 gm (J-235/gm-mole
- i. e p g O.85 The term, ag /ag , accounts for the fact that not all U-235 atoms which absorb neutrons are fissioned.
i 4 a 23.8 6 23 FD = X .02 X 10
- f. ;.
avs W 235 X 0.85 g! 20
= T.46 X 10 fissiom/cc .
;I "
~ - - maz:nw __ _ - _ _ - - - ~ -n--= - - ~ ~ -
=- - -
._ . x
,5=~ N e ; + ~: ?.%ri : .A-- x.= . . _.. -l:=-uik;i-0.ih,bpk ,;h ;Ue&, n, . . . -- - e m .
,; - 7.~p;' - . _ _ _
+
4 i '
\
. .]
._ j
~ s i
.i -
Il [ Rev 4/78 Peck to average flux ratio in the core is opproximately 1.86 3 - FD (A9) pk = 1.86FDavg 0
! = 2.72 X 10 fissions /cc ).j w jl Control rod fuel elements are 35% depleted at end of life. Similar fission l c density calculations yield }e_i.
W' Rev 4/78 FD 207 ;,,;,,,7 ,, 3 pk = 5.44 X 10 , a m. [ , A.4 FISSILE MATERIAL DENSITY a s.. r' t t vr atu]rne.-Fissi c material density in fuel pl<stes is the - % plate loading divided C i s'
' ~ W235 -
, FMD = y .-
, (A10)
.4, 140gm
'~
,i , = .-
ly . ;" 354 cc
~
I
= .395 gm/cc 2#
A.5 s REACTOR HEAT FLUX _ s
@~ . -!" MW s-Q =
(All)
-- e mg A
- s.
1 - % _ j where MW=2MW 6
.1 t '
= 2 X 10 watts
= 6 6.82 X 10 BTWhr J
6 , wy -s Q -- 6.82 X 10 s
-! , avg 345 i , 1 c, c
= 1.98 X,10' BTWhbft ;
j Peak hoot flux corresponds to peak neutron flux. N ,
;i Rev 4/78 -T (A12) s , Qpk = 1.86 Qavg 1,
2
, -= 3.68 X 10' BTU /hr-ft
- ] 1 a' ,
11 ! y.) - s
;} .
- i .
P S g . Ad e. h a 44 %dD l't ., pA U. * * . . N e " r_ Y ep-' " _b. I " ,,9
- - e -
;^- ~ 7zw &;. ::h.b?::22t51?lli252?l"b " Xi-:
'.'., . a_ ;. . .
L1 d * ' g y n
"I A.6 COOLANT VELOCITY 3
4 p j. d . v=7C (A13)
?) -
3 where F = 980 gpm coolant flow rate
]j = 2.19 ft3/sec S1 .
4 i M 2.19 nj
.( ii i
.884
=
= 2.48 ft/sec
= 8919 ft/hr A.7 REYNOLD'S NUMBER
- i The average coolant temperature in the core is assumed to be 120 F. The 1, average temperature near the clad - coolant interface is assumed to be 135 F, the mean between 120 F and on estimated 150 F clad temperature.12 135 F is the tempe.w; ore used to evaluate fluid parameters, i
[ Re = "P * (A14) P 1 i' where - v = 8919 ft/hr coolant velocity 4 p= 61.5 lb/ft3 coolant density p = 1.19 lb/fdhr coolant viscosity j l = Characteristic channel dimension, 2r = 0.0196 ft c l
, I was chosen as .0196 feet based upon flow analysis between parallel plates.14 The characteristic channel dimension is the hydraulic diameter of the channel , which is equal to four times the cross sectional crea divided by the wetted perimeter.
a 4 Re = 8919 X 61.5 X .0196 1.19 4
- . Rev 1Q/77 9.05 X 10
.. For a smooth channel of assumed relative roughness, G/D = 5 X 10 -5 , the ti Reynold's Number describes flow in the transition region from laminar to 'I turbulent flow on the Moody Friction Chart, Figure A1. c r .9 ', A.8 PRANDTL NUMBER n ] . h:] l Pr= [pC (A15)
.3 i
i: ._ !4 18 . _a; ,- g _ - % % _ _ - . - - - - - * - -- -
~
~ ;r.. . ,. . . .
-~ v n .i.:: 2 2;,t.n.s t~ n a m ;h w a:1;- < . ::.'. .';' - . . C T T^:;'"~~~'=., .' .L
.:dk.. . - . . . . . . ' . ~ .-
4 . .
- 1-(- .
$.f ' a . Aj ' where c = .998 BTLKIb- F coolant specific heat
@ 18 = .377 BTLVhr-ft- F coolant thermal resistance d9 1., I, Pr = i .19 X .998 ti .
.377
'l 2 = 3.15 a, .
j A.9 NUSSELT NUMBER q N V 3 The Dittus - Bolter equation is used to evaluate Nusselt Number. SI
~!.i
. Nu = 0.023(Re)0.8(Pr)0.4 (A16) yA
!ji Rev ly77 = 53 0.
A.10 HEAT TRANSFER COEFFICIENT
~
i _', j 1 t,} 5 1 The heat transfer coefficient for the film at the clad - coolant interface -
] ,, is j
h= (A17) i
'{. 1^ = .377 X 53 '
- i .0196 2
;'] Rev ly77 =1022 BTI.Vhr-ft - F
- 1 ij A.11 THERMAL RESISTANCE
)
- sj A model of a coolant channel is shown in Figure!A2. The value of thermal
. ] "' resistivity, K c, for the clad is the value for 1100F aluminum. The value
. }i For K is an experi. mentally detenr.ined vah.e for oxide cermets similar to the p,owder metallurgy aluminides and oxides that will be used in the new i FNR fuel.10 <
> 1
. A.12 MAXIMUM FUEL Pl ATE CLAD TEMPERATURE l
.f 1 From Figure A2, the differential tempemture between the moor centerline, l t , and bulk coolant, t' , is evaluated by 1' m W
'. : t
-t*
JJ , i Qpk = R +R+R_ (A18) e 7 j4 m > . .e 4 since R_ >> 'R and R .
- J." ! j F ~ c m
,I
.J ,'
;I j! , -
~
e . g i '
.~ n : _.n._ . _, . _ . , . _ _ , ._ __
--- n. ,, e m-.m; m m - m .,
+
+
t ., p k a s
~t e
L p i m I w ,
- n. Fr 1
. p E ((-
; y ssouybnoJ aApolad O -
Z ' i-g c @@ T N $
~
/O O '
OO Q O O 688 8 8 boo dO O OO 00O O O 0o00 O O O O O
- dg-h ...
- " ?- a O
t.
)
'e' 8 "-
U O ,
! , ,', ,/
M N
, ' f,N .
=
sD M
,/ P' s*9 D ,-
i' s o I i Air 111 - J
,o -
A Q M
~
( .
! ! f jf ,f.f =*e m
- in 1
i i a f. O J $ '
' b Af -
u o m V
. I s ts m i ' *
'e 1lA g ,2 W
- l. f ,f,g e- z
~'~ ; : ,
, ,, - i:; :
O 2 AC M T _. k, e e
, --= .. . .i .l .I ,i ,1,/,1 - f..
x - Ill l r fH -g a L y
' G..
k' -
? .
I Illl/F E ! E 9 - 8I i w .
/ t ,! ,I,I,U,,f, .
si
-e ~k i -
t i li i i H//#/ \ - D U W I ll l l i /dff L :
- b
--L / Jllll /f i i~ J .
- P; f/ /N,viinp N 1
<s m O L
e
-*9 t-
= , 1 i z i 1 -
.I I I I r //N/f M A "
g l .* Ii I / /JH/f/ : s- <C Y
' # 4 ///S.'<7 -
- Z >
w S.I
.- /
3 -
- u s.#
r l ~j2( , n op . 5d u. O [ - 9 y,
~ .o ~ r L.Oe L M.s i
.........I,.. ,,,, ,,,, ,,, ,,,,, , .:. W O
r' N Omm~m o n n m n ..
-000 0 0 3 0 ~ O -
v
. cocc o c o d o o o
O d OO8 c0 ty; t e,r s co Z p
*ftp)
- ,- O ..
sy = / Jotso; uogsp3
- t .I aC j' '
"- l pI g.
E o h. tl:l
.f5 .-
h r L p
)
. . +1
, c en g,- mt.w ?- e .w - -
- v7 7. . ~ .1 .- :v ., . w.....r.7'..,.
^' gu. fl
- a. *. . r : .", . .s . . . -,*,,',,,,,,,.,_y.w.
.. ..L- .; 1%.. ~ .+,4 x , n . _ , ...._..,._,_1 i , ' , ' /, (;fn .l
h e u._.,.__.yg.===_,.-.=b_.
. m _. - 2 ,_ _ a ,2. _ ,.._.3.,_ .... c . . . .
.m,j.,,.,,._.#_.__.,, n _ .. _
6.1 q . .
. i
{ - 1 ik If ..; me AT- C t. A o PILM m {* ; a - 6X 3 : : 6X c > j f b Emc. b BdLR l Cool A NT a 1 1
,[b=12o*F
.s a
.cz
* .o :
5 4, tuc. fCe Eta Rm Rc Rs 9 *! k m
.00083 -6 2 p) 4 R
m
=
K .
=
103
= 8 X 10 hr-ft o F/ BTU .- m h
9 .. Mc
' 8 t :I
*00167 o ,
; i R=K c
=
115
= 1.4 X 10-5 hr-ft2o F/ BTU d4 C <1 4 -IkI 2 Rev 10/77 Rp = I/h = 1/1022= 9.8 X 10 hr-ft - F/ BTU l M.
l'i i I
.:'1 g!
7.1 m 2 . 04 i 6i
+.
FIGURE A2 ij ' FNR FUEL THERMAL RESISTANCE MODEL M l
~
v w! i. . 3 .i . s '~ - . t .. . _ _T _ % aw *we > i.wdA. - '
- - ^" "" "- ' "
3
. . . ~ - . . . - ,
~ x
~. a, .o. . .-
..w --- .
w2; . ,. ..-= au rw'w = .=>
" -s:
- =. .
.g-
- 4, n . . .
T3 :- s] ' M~ t -t CW W f! e Q (A19) P k= R p a Q ad
't S
p t =t .,s cw m u . M1
,l which means the temperature rises in the clad and meat are negligible compared > to the rise across to clad - coolant interface. The fuel element plates are 5 essentially at a uniform radial temperature, tcw . From equation (A19),
y t =t w
+- Q R1-l cw pk' ~.
0 4 -4 o y Rev 4/78 = 120 +3.68 X 10 X 9.8 X 10 F 6 .
. = 120 + 36 A.
Il i
= 156 F j
i
~
h a e 3 k u.>
-~.J.
7: e i. l'l . f
'.: t .
Y 3 1 y n-(i ?i k i 3
- < t
!j:, ;i . ,g + t ,. ,I ; r
*M % m.
- - - -__ m_ __ _
a _:- ._ _ -- . .w_=- ; . /~r: . .yk;.Gs .. ~_._. ~ % if'-N_Y-
;. . ~ . ~
b O
- d
? J' .
di REFERENCES 5 .I - N
- 1
.g j 1. G. W. Gibson, "The Development of Powdered Uranium Aluminide Compounds y; for Use as Nuclear Reactor Fuels", IN-1133, TID-4500, December,1967 'l i 2. M. M. Martin, A. E. Richt, and W. R. Martin, " Irradiation Behavior of il Aluminum Base Fuel Dispersions", ORNL-4856, May,1973.
7j 3. " Reactor Safety Evaluation of ORNL Proposal to Modify Fuel in ORR", ORNL,
,@ .j February,1977.
j 4. P. Tichler, " Review of Proposed increase in Fuel Element Loading and Fuel j ' Burnup", memorandum, Brookhaven National I.aboratory, February,1977 11 5. W. C. Francis, " Fuel Elements for Thermal Test Reactors - Performance at
]; NRTS", Paper for presentation at the AEC - Industry Meeting, Water Reactor '4 . ; -
Fuel Element Technology, U. S. AEC Headquarters, Germantown, Maryland, January 29-30, 1968. Jj] '6. C. Julian, " Evaluation of a 6.2 Kilogram U-235 Core Loading for the Missouri
,j University Research Reactor", Docket 50-186, Technical Specification Request jq for Change, University of Missouri, July 28,1970.
j 7. J. H. Crawfont and M. C. Wittels, " Radiation Stability of Nonmetals and j , Ceramics", Proceedings of the Second United Nations International Conference
! on the Peaceful Uses of Atomic Energy, Geneva,1958, Vol. 5, United Nations, l 'l Geneva,1958, pp. 300-310.
t j* 8*.
~
G. W. Gibson and O. K. Shupe, Annual Progress Report on Fuel Element jj Development for FY 1961, U. 5. Atomic Energy Commission Report, 10 0-16727, J March,1962.
.t 1 9. W. J. Werner and J. R. Barkman, " Characterization and Production of U 0 l] for the High Flux Isotope Reactor", ORNL-4052, April,1967. 38
'1
, ;. 10. D. L. McElroy, R. S. Graves, and J. P. Moore, " Physical Properties of g .] Two-Phase Materials Used in Fuel Cores", Metals and Ceramics Division Annual y:j Progress Report, ORNL-4570, Octobe,1970.
.j l --
11; Reactor Engineering Branch Annual Report, IN-1335, Idaho Nuclear Corporation,
'j November,1969.
~
[{ 12. A. J. Chapman, Heat Transfer, McMillan Publishing Co., New York,1974. j j;j 13. L. F. Moody, " Friction Factors for Pipe Flow", Transactions ASME, Vol. 66, No. 8,1944. l j 14. M. M. El-Wakil, Nuclear Heat Transport, Intext Educational Publishers, 1 New York,1971, p. 248. 9' I 15. W. Dienst, S. Nazare, and F. Thummler, "Irradiction Behavior of UAlxDispersion jj q . Fuels forThermal High Flux Reactors", Journal of Nuclear Materials, 64, 1977 . '1. y3 ]$ - 1, !,
. - . ~.rwm_ ._ .. :--- n - w -- .-___-. - - -~- --
i
. .r -
=Lw am.g=: - . .. ,. % :;.3-, 3:7 .:3g.,gg; g73, 3 ,_.3 ;..-- &-1;;+g
~.> s.
-4 ; . a j ... -
'4 i.
-37 .
.i :
lij ! 16 M. Graber and R. Hobbins, "Irrodiotion Testing of Sample Fuel Plates to Very
?i i High Burnups", INC-16-1, U. S. Atomic Energy Commission Report ANCR-1016, 4: October,1971. .. ;s ;
jj j
- 17. R. Hobbins, "Irrediation Experiment", INC-16-2, Aeroiet Nuclear Company,1974.
@. : 18. M. Graber, " Radiation Effects on Dispersion Fuels", Annual Progress Report n i on Reactor Fuels and Materials Development, U. S. Atomic Energy Commission Report 10 0-17154, February, .1966 /3;
- e. .
=,
. b. e n i d?
?s 15 {4
- a. .
I ;* s
't
'A l 6
1
,1, -
3 .
, 1 8 e
1 1 l
.} * *$ 6 8 .( .' r.
e e r..' . I as 1' di
's
'r,
~
3 m$* h L* f *h., s, .)
.'i s
- 4 5
-u I b<i : '
- r. . !
3: h, I
.:l.; -
n-,, , -- , -- . , . , , ,..0 --
. . - _ :.}}