coal power
DESCRIPTION
HydrocarbonTRANSCRIPT
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CHAPTER 6
CHEMICAL PROCESSING*
6-l . INTRODUCTION
One of the prncpal advantages of flud fuel reactors s the possblty of
contnually processng the fuel and blanket materal for the removal of
fsson products and other posons and the recovery of fssonable materal
produced. Such contnuous processng accomplshes several desrable
objectves : (a) mprovement of the neutron economy suffcently that the
reactor breeds more fssonable materal than t consumes, (b) mnmza-
ton of the hazards assocated wth the operaton of the reactor by man-
tanng a low concentraton of radoactve materaln the fuel, and (c) m-
provement of the lfe of equpment and stablty of the fuel solutonby
removng deleterousfsson and corroson products .
The performance
and operablty of a homogeneous reactor are consderably more dependent
on the processng cycle than are those of a soldfuel reactor, although the
objectves of processngare smlar .
The neutron posonngn a homogeneous reactor from whch fsson
product gases are removed contnuously s largely due to rare earths[1],
as shown n Fg . 6-1 . In Fg . 6-1 the rare earths contrbutng to reactor
posonng are dvded nto two groups .The tme-dependent rare earths
are those of hgh yeld and ntermedatecross secton, such as Nd
143
and N d145 Pr141
and Pm 147 , whch over a perod of several months could
accumulate n the reactorand result n a posonng of about 2070 . The
constant rare-earth poson fractons due prmarly to
Sm149 and Sm
ls1 ,
whch have very largecross sectons for neutron
absorpton but low yeld,
and therefore reach ther equlbrum leveln only a few days' operaton .
Posonng due to corroson of the stanless-steel reactor system was cal-
culated for a typcal reactor contanng 15,500 ft2 of steel corrodng at a
rate of 1 mpy . It s assumed that only the nckel and manganese contrbute
to the posonng, snce ron and chromum wll hydrolyze and precptate
and be removed from the reactor system; otherwse, corroson product
posonng would be four tmes greater than ndcated n Fg . 6-1 . The
control of rare earthsand corroson product elements s dscussed n sub-
sequent sectons of ths chapter .Removal of solds from the fuel soluton
also mproves the performance of the reactor by dmnshng the deposton
of scale on heat-transfer surfaces and reducng the possblty of eroson of
pump mpellers, bearng surfaces, and valve seats .
*By R. A. ItlcNees, wth contrbutonsfrom W. E. Brownng, W. D. Burch,
R. E. Leuze, W . T. 11cDuffee,and S. Peterson, Oak Rdge Natonal Laboratory .
-
FIG. 6-1 .
Poson effect as a functon of chemcal group n core of two-regon
thermal breeder .
FIG. 6-2 .
Conceptual flow dagram for processng fuel and blanket materal from
a two-regon reactor .
The bologcal hazards assocated wth a homogeneous reactor are due
chefly to the radoactve rare earths, alkalne earths, and odne [2] .
The mportance, as a bologcal hazard, of any one of these groups or nu-
cldes wthn the group depends on assumptons made n descrbng ex-
posure condtons ; however, P 31 contrbutes a major fracton of the rada-
ton hazards for any set of condtons . Whle the accumulaton of hazardous
materals such as rare earths and alkalne earths wll be controlled by the
processng methods to be descrbed, less s known about the chemstry of
-
FG. 6-3 .
Conceptual flow dagram for processng blanket materal from a two-
regon plutonum producer .
odne n the fuel systems and methods for removng t. Exstng nforma-
ton on odne processng s dscussed n Secton 6-5.
Schematc flowsheets for proposed processng schemes for two types of
two-regon aqueous homogeneous reactors are shown n Fgs. 6-2 and 6-3 .
In both cases, solds are removed by hydroclones and concentrated nto
a small volume of soluton for further processng .The nature of such
processng wll be determned by the exact desgn and purpose of the
reactor. Thus, for a two-regon plutonum producer, the core and blanket
materals would have to be processed separately to avod sotopc dluton,
whle for a thorum breeder, core and blanket materal could be processed
together. However, f an attractve method should be developed for leach-
ng uranum and/or protactnum from a thorum-oxde slurry wthout
serously alterng the physcal propertes of the slurry, the two materals
could be processed separately .In a smlar way, the relaton between
odne control and fsson product gas dsposal s such that nether problem
can be dsassocated from the other . A specfc, complete, and feasble
chemcal processng scheme cannot be proposed for any reactor wthout
an ntmate knowledge of all aspects of desgn and operaton of the reactor.
However, some of the basc chemcal knowledge needed to evaluate varous
-
possble processng methods has been developed and s presented n the
followng sectons .
6-2 . CORE PROCESSING : SOLIDS
REMOVAL
6-2
.1 Introducton. Early n the study of the behavor of fsson and
corroson products n uranyl sulfate solutons at temperatures n the range
250 to 325 C, t was found that many of these elements had only a lmted
solublty under reactor condtons . Detaled studes of these elements
were conducted and devces for separatng solds from lqud at hgh
temperature and pressure were constructed and evaluated . Based on ths
work, a plot plant to test a processng concept based on solds separaton at
reactor temperature was nstalled as an adjunct to the HRE-2 . These
processng developments are dscussed n ths secton .
6-2
.2 Chemstry of nsoluble fsson and corroson products . Of the
nongaseous fsson products, the rare earths contrbute the largest amount
of neutron poson to a homogeneous reactor after a short perod of operaton
(Fg . 6-1) .
Therefore, a detaled study of the behavor of these elements
has been made. All the rare earths and yttrum showed a negatve tem-
perature coeffcent of solublty n all the solutons studed and a strong
tendency to supersaturate the solutons, as shown n Table 6-1 . Wth the
excepton of praseodymum and neodymum, whch are reversed, the solu-
blty at a gven temperature and uranyl sulfate concentraton ncreased
wth ncreasng atomc number, wth yttrum fallng between neodymum
TABLE 6- 1
SOLUBILITY OF
LANTHANUM SULFATE IN
0 . 02 m U02SO4
SOLUTION
0 . 005m H2SO4 AS A
TEMPERATURE
FUNCTION OF
mg La2(S04)3/kg H2O
Temperature,
CTrue
solublty
Concentraton
ntate precptaton
requred to
190 250
760
210 130 360
230 54 167
250
25 77
270
12 36
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and samarum, as shown n Table 6-2 . Increasng the uranyl sulfate
concentraton ncreased the solublty of a gven rare-earth sulfate, as
shown n Table G-3 .
In a mxture of rare-earth sulfates the solublty of an ndvdual rare
earth s less than t would be f t were present alone. For example, the
solublty of praseodymum sulfate at 280 C s 170 mg/ kg H 2O wth no
other rare earths present, as compared wth 12 mg, kg H 2O n a soluton
made up wth a rare-earth mxture contanng ( ; / praseodymum sulfate .
Samples of the precptatng salts solated from soluton at 280 C have
usually been the sulfates and contaned no uranum . However, under
specal condtons a mxed sulfate salt of neodymum and uranum has been
observed [3] .
The alkalne earths, barum and strontum, also show a negatve tem-
perature coeffcent, but not so strongly as do the rare earths; almost no
effect can be seen when the temperature of precptatng solutons s n-
TABLE
6-3
EFFECT OF URANYL SULFATE CONCENTRATION ON THE SOLUBILITY OF
NEODYMIUM SULFATE AT VARIOUS TEMPERATURES
Nd2(S04)3 solublty, mg/kg H20
U, g/kg H20
250 C280 C 300C
5 .7
270 115 73
10 .8
400 200 120
16 .6 770 300 180
22 .4 > 1000 500 300
SOLUBILITY OF VARIOUS
0 .02 mU 02S04
TABLE 6-2
RARE-EARTH SULFATES
0 .005m H2SO4 AT 280 C
IN
Salt
mg/kg
Solublty,
H20
Salt
mg/kg
Solublty,
H2O
La2(S04)3 10 Nd2(SO4 ) 3 110
Ce2(S04)3 50 Y2(S04)3 240
Pr 2(S0 4 ) 3 170 Sm2(S04)3 420
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creased from 250 to 300 C. At 295C n 0.02 m U02S04-{1.005 m H2SO4
soluton, the solublty of barum sulfate s 7 mg/kg H2O and that of
strontum sulfate s 21 mg/kg H20 . Both the alkalne and rare-earth
sulfates show a strong tendency to precptate on and adhere to steel
surfaces hotter than the precptatng solutons, and ths property can be
used to solate these solds from lquds at hgh temperatures .
Other fsson and corroson product elements hydrolyze extensvely at
250 to 300C and precptate as oxdes, leavng very low concentratons
n soluton . Iron(III) at 285C has a solublty of 0 .5 to 2 mg Fe/kg H 2O
and chromum(III), 2 to 5 mg/kg H20 . At 285C less than 5 mg of zr-
conum or nobum per klogram of H2
O remans n soluton .
For other elements of varable valence, such as technetum, the amount
of the element n soluton s determned by the stable valence state under
reactor condtons . In general, the hgher valence states better resst hy-
drolyss and reman n soluton . Thus at 275 C n 0.02 m U02SO4 Te(VII)
s reduced to Tc(IV) f hydrogen s present, and only 12 mg/kg H2O re-
mans n soluton . However, a slurry of Tc02 n the same soluton but wth
oxygen present dssolves to gve a soluton at 275 C wth a technetum
concentraton of more than 9 g/kg H20. The same qualtatve behavor s
observed wth ruthenum . Selenum and tellurum n the hexapostve state
are much more soluble than when n the tetrapostve state [4] .
A few elements, e.g ., cesum, rubdum, nckel, and manganese, ntro-
duced nto the fuel soluton by fsson or by corroson of the system, are
very soluble under reactor condtons . Ther removal and control are ds-
cussed n Secton 6-4 .
6-2
.3 Expermental study of hydroclone performance . It s evdent
from the precedng secton that the amount of uranum wthdrawn from
the reactor dmnshes f the collecton, concentraton, and solaton of the
nsolubles can be effected at hgh temperature . One devce capable of
collectng and concentratng solds at hgh temperature s a sold-lqud
cyclone separator called a "hydroclone," or "clone ." A dagram of a hydro-
clone s shown n Fg . 6-4. In operaton, a solds-bearng stream of lqud
s njected tangentally nto the wde porton of a concal vessel . Solds
concentrate n a downward-movng layer of lqud and are dscharged from
the bottom of the clone nto the underflow recever . Partally clarfed
lqud leaves from the top of the clone through a vortex fnder. Use of the
underflow recever elmnates mechancal control of the dscharge flow
rate and, by proper choce of hydroclone dmensons, any desred rato of
overflow rate to underflow rate can be acheved . The drvng force for the
system s provded by a mechancal pump .
The factors nfluencng the desgn of an effectve hydroclone for homo-
geneous reactor processng use have been studed, and hydroclone desgns
-
FIG. 6-4 .
Schematc dagram of a hydroclone wth assocated underflow recever .
based ol these studes have been tested n the laboratory and on varous
crculatng loops [5] .All tests have shown conclusvely that such hydro-
clones can separate nsoluble sulfates or hydrolyzed materals from lqud
streams at 250 to 300 C. In the HRE-2 mockup loop a mxture of the sul-
fates of ron, zrconum, and varous rare earths, dssolved n uranyl-
sulfate soluton at room temperature, precptated when njected nto the
loop soluton at 250 to 300 C. The solds concentrated nto the underflow
recever of a hydroclone contaned 75% of the precptated rare-earth
sulfates . When the lanthanum-sulfate solublty n the loop soluton was
exceeded by 10 U%, the concentraton of rare earths n the underflow recever
was four to sx tmes greater than n the rest of the loop system; some
accumulaton of rare earths was observed n the loop heater. A large
fracton of the hydrolyzed ron and zrconum was collected n the gas
separator porton of the loop . III the separator the centrfugal moton
gven to the lqud forced solds to the perphery of the ppe and allowed
them to accumulate . Only about 10`/, of the solds formed n the loop was
recovered by the hydroclone, and examnaton of the loop system ds-
closed large quanttes of solds settled n every horzontal run of ppe.
-
FIG . 6-5
. Schematc flow dagram for the HRE-2 chemcal processng plant .
TABLE
6-4
DIMENSIONS OF HRE-
2 HYDROCLONES
Symbol Locaton
Dmenson, n .
0 .25-n .
hydroclone
0.40-n.
hydroclone
0
.56-n .
hydroclone
DL Maxmum nsde
dameter
0 .25 0 .40 0 .56
L
Du
Insde length
Underflow port
1
.50
2 .40 3 .20
Do
dameter
Overflow port
0 .070 0 .100
0 .148
DF
dameter
Feed port effec-
0 .053 0 .100 0 .140
tve dameter 0 .0510 .118 0 .159
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Samples taken from the loop after addton of preformed solds and wthout
the hydroclone operatng showed an exponental decrease n solds concen-
traton wth a half-tme of 2 .5 hr; wth the hydroclone operatng, the half-
tme was 1 .2 hr. In the HRE-2 chemcal plant [5], operated wth an aux-
lary loop to provde a slurry of preformed solds n uranyl sulfate soluton
as a feed for the plant, the half-tmes for solds dsappearance and removal
were 11 hr wthout the hydroclone and 1 .5 hr wth t. The effcency of the
hydroclone for separatng the partcular solds used n these experments
way about 10,' ;, . Wth gross amounts of solds n the system, concentraton
factors have been as large as 1700 .
Correlaton of these data wth antcpated reactor chemcal plant oper-
atng condtons ndcates that the HRE-2 chemcal plant wll hold the
amount of solds n the fuel soluton to between 10 and 100 ppm. If neces-
sary. performance can be mproved by ncreasng the flow through the
chemcal plant and by elmnatng, wherever possble, long runs of hor-
zontal ppe wth low lqud velocty and other stagnant areas whch serve
to accumulate solds .
6-2.4 HRE-2 chemcal processng plant.* An expermental faclty to
test the solds-removal processng concept has been constructed n a cell
adjacent to the HRF-2. A schematc flowsheet for ths faclty s shown
n Fg. 6-5 .
A 0.75-gpm bypass stream from the reactor fuel system at 280 C and
1700 ps s crculated through the hgh-pressure system, consstng of a
heater to make up heat losses, a screen to protect the hydroclone from
pluggng, the hydroclone wth underflow recever, and a canned-rotor
crculatng pump to make up pressure losses across the system . The
hydroclone s operated wth an underflow recever whch s draned after
each week of operaton, at whch tme the processng plant s solated
from the reactor system .
At the concluson of each operatng perod 10 lters of the slurry n the
underflow pot s removed and sampled. The heavy water s evaporated
and recovered, and the solds are dssolved n sulfurc acd and sampled
agan. The soluton s then transferred under pressure to one of two 100-gal
decay storage tanks. Followng a three-month decay perod, the soluton
s transferred to a shelded carrer outsde the cell and then to an exstng
>olvent extracton plant at Oak Rdge Natonal Laboratory for uranum
decontamnaton and recovery . The sulfurc acd soluton step s ncor-
pono(d n the 1-111,1;-2 chemcal plant to ensure obtanng a satsfactory
-ample . Ths step would presumably not be necessary n a large-scale
plant .
*Contrbuton from 11' . D . Burch .
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FIG. 6-6 . HRE-2 chemcal plant cell wth equpment.
-
All equpment s located n a 12- by 24- by 21-ft underground cell located
adjacent to the reactor cell and separated from t by 4 ft of hgh-densty
concrete . Other constructon features are smlar to those of the reactor
cell, wth provsons for floodng the cell durng mantenance perods n
order to use water as sheldng. Fgure 6-6, a photograph of the cell pror
to nstallaton of the roof plugs, shows the maze of ppng necesstated by
the expermental nature of ths plant .
Dmensons of the three szes of hydroclones desgned for testng n ths
plant are shown n Table 6-4 . These three hydroclones, whch have been
FIG. 6-7. HRE-2 chemcal plant hydroclone contaner .
selected to handle the range of possble partcle szes, are nterchangeable
at any tme durng radoactve operaton through a unque, specally ma-
chned flange, shown n Fg. 6-7 . Removal of the blnd closure flange ex-
poses a cap plug and retaner plug . Removal of these wth long-handled
socket wrenches permts access to the hydroclone tself . Ths operaton has
been performed routnely durng testng wth nonradoactve solutons .
In processng homogeneous reactor fuel, a transton from a heavy- to a
natural-water system s desrable f fnal processng s to be performed n
conventonal solvent extracton equpment . Such a transton must be ac-
complshed wth a mnmum loss of D20 and a mnmum contamnaton of
-
the fuel soluton by H2
O n recycled fuel. Intal tests of ths step n the
fuel processng cycle have been carred out [6]. In these experments a
mxture of 57 D20, 957 H 2O was used to smulate reactor fuel lqud.
The dssolver system was cycled three tmes between ths lqud and or-
dnary water, wth samples beng taken durng each porton of each cycle .
Isotopc analyss of these samples showed no dluton of the D20 n the
enrched soluton and no loss of D20 to the ordnary water system .
At expected corroson rates, approxmately 400 g of corroson products
wll be formed n the reactor system per week, and the underflow recever
was therefore desgned to handle ths quantty of solds. The adequacy of
the desgn was shown when more than three tmes ths quantty of solds
was charged to the underflow recever and draned n the normal way wth-
out dffculty .
Full-scale dssoluton procedures have also been tested[6] . To mnmze
the possbltes of contamnatng the reactor fuel soluton by foregn ons,
a dssoluton procedure was developed usng only sulfurc acd . Ths con-
ssts of a 4-hr reflux wth 10.8
M
H2S04 n a tantalum-lned dssolver fol-
lowed by a 4-hr reflux wth 4 AI H2SO4, and repeated as requred untl
dssoluton s complete . Decay storage tanks and other equpment requred
to handle the bolng 4 M H2S04 are fabrcated of Carpenter-20 stanless
steel. Tests have repeatedly demonstrated more than 99 .57 dssoluton
of smulated corroson and fsson products n two such cycles .
The HRE-2 hydroclone system has been operated as an ntegral part
of the reactor system for approxmately 600 hr and for an addtonal
1200 hr wth a temporary pump loop durng ntal solds-removal tests .
Durng ths operatng perod, n whch smulated nonradoactve fuel
solutons were used, the performance of the plant was satsfactory n all
respects .
6-3. Fsson PRODUCT GAS DISPOSAL*
6-3.1 Introducton . To prevent the polluton of the atmosphere by
radoactve krypton and xenon sotopes released from the fuel soluton, a
system of contanment must be provded untl radoactve decay has re-
duced ther actvty level. Ths s accomplshed by a method based on the
process of physcal adsorpton on sold adsorber materals . If the adsorber
system s adequately desgned, the ssung gas stream wll be composed of
long-lved Krs 5, oxygen, nert krypton sotopes, nert xenon sotopes, and
nsgnfcant amounts of other radoactve krypton and xenon sotopes . In
case the actvty of the KrS 5s too hgh for dluton wth ar and dscharge
to the atmosphere, the mxture may be stored after removal of the oxygen
*Contrbuton from W . E. Brownng .
-
Fc .6-8. Adsorpton of krypton on varous adsorbents at 28 C .
or further separated by conventonal methods nto an nert xenon fracton
and a fracton contanng Kr8 ' and nert krypton .
6-3 .2 Expermental study of adsorpton of fsson product gases . Evalu-
aton of varous adsorber materals based on expermental measurements of
the equlbrum adsorpton of krypton or xenon under statc condtons s
n progress [7] . Results n the form of adsorpton sotherms of varous
sold adsorber materals are presented n Fg . 6-8 .
A radoactve-tracer technque was developed to study the adsorpton
effcency (holdup tme) of small, dynamc, laboratory-scale adsorber
systems [8] . Ths conssts of sweepng a bref pulse of Kr 83 through an ex-
permental adsorber system wth a dluent gas such as oxygen or ntrogen
-
and montorng the effluent gases for Kr85 beta actvty. The actvty n
the gas stream versus tme after njecton of the pulse of Kr 85 s recorded .
A plot of the data gves an expermental eluton curve, such as shown n
Fg
. 6-9, from whch varous propertes of an adsorber materal and ad-
sorber system may be evaluated .
Among the factors whch nfluence the adsorpton of fsson product
gases from a dynamc system are (1) adsorptve capacty of adsorber ma-
teral, (2) temperature of adsorber materal, (3) volume flow rate of gas
stream, (4) adsorbed mosture content of adsorber materal, (5) compos-
ton and mosture content of gas stream, (6) geometry of adsorber system,
and (7) partcle sze of adsorber materal . The average tme requred for
the fsson product gas to pass through an adsorber system, tmax, s nflu-
enced by the frst fve of the above factors . The shape 'of the expermental
eluton curve s affected by the last two .
The temperature of the adsorber materal s of prme mportance . The
lower the temperature the greater wll be the adsorpton of the fsson gases,
and therefore longer holdup tmes per unt mass of adsorber materal wll
result. The dependence of adsorptve capacty, k, on temperature asde-
termned by holdup tests wth some sold adsorber materals s shown n
Table 6-5 .
TABLE
6-5
ADSORPTIVE CAPACITY OF VARIOUS MATERIALS AS A FUNCTION
OF TEMPERATURE
Gas Dluent Adsorber
cc gas/g
273K
adsorbent*
323K 373K
Xe
Kr
Kr
Kr
Kr
02
He
02
orN2
02
02
Charcoal
Charcoal
Charcoal,
Lnde Molecular
Seve 5A
Lnde Molecular
Seve lox
4 .7 X 103
1 .8 X 102
68
23
11
4 .0 X 10 2
34
24
9
5 .7
80 .0
9 .6
11 .0
4 .5
3 .5
*Gas volume measured at temperatures ndcated .
-
FIG. 6-9 . Expermental Kr85 eluton curve .
At a gven temperature, the average holdup tme,tmax,
s nversely pro-
portonal to the volume flow rate of the gas stream . If the volume flow
rate s doubled, the holdup tme wll be decreased by a factor of two .
All the sold adsorher materals adsorb mosture to some degree . Any
adsorbed mosture reduces the actve surface area avalable to the fsson
gases and thus reduces the average holdup tme .
The geometry of the adsorber system nfluences the relaton between
breakthrough tme,tb,
and average holdup tme, t max , as shown n Fg . 6-9 .
Ideally, for fsson product gas dsposal, a partcular atom of fsson gas
should not emerge from the adsorber system pror to the tme t,,, ax . Snce
ths condton cannot be realzed n practce, the dfference between break-
through and average holdup tmes should be made as small as possble .
For a gven mass of adsorber materal a system composed of long, small-
dameter ppes wll have a small dfference betweentb
and t,nax, whereas a
system composed of short, large-dameter ppes wll not .
The partcle sze of the adsorber materal s mportant for ensurng n-
tmate contact between the actve surface of the adsorber materal and
the fsson gases . A system flled wth large partcles wll allow some mole-
-
cules of fsson gases to penetrate deeper nto the system before contact s
made wth an actve surface, whle the pressure drop across a long trap
flled wth small partcles may be excessve. Materal between 8 and 14
mesh n sze s satsfactory from both vewponts.
6-3 .3 Desgn of a fsson product gas adsorber system . The desgn of
an adsorber system wll be determned partly by the fnal dsposton of
the effluent gas mxture . If ultmate dsposal s to be to the atmosphere,
the adsorber system should be desgned to dscharge only Kr 8' plus nert
krypton and xenon sotopes. If the effluent gases are to be contaned and
stored, the adsorber system may be desgned to allow dscharge of other
radoactve krypton and xenon sotopes . In the followng dscusson t s
assumed that fnal dsposal of the effluent gas mxture wll be to the at-
mosphere. The followng smple relaton has been developed whch s use-
ful n fndng the mass of adsorber materal n such an adsorber system :
M = tmax,
where A1= mass of adsorber materal (grams), F = gas volume flow rate
through adsorber system (cc/mn), k = adsorptve capacty under dy-
namc condtons (cc/g), andtmax
= average holdup tme (mn) .
The operatng characterstcs of the reactor wll dctate the composton
and volume flow rate of the gas stream ; tmn x wll be determned by the al-
lowable concentraton of radoactvty n the effluent gas ; k values for
krypton and xenon must be determned expermentally under condtons
smulatng these n the full-scale adsorber system . It should be noted
(Fg. 6-9) that a porton of the fsson gas wll emerge from the adsorber
system at a tme th pror to the average holdup tme,tmax .
The desgn
should ensure that radoactve gas emergng at tmeto
has decayed suff-
cently that only nsgnfcant amounts of actvty other than KrS 5 wll be
dscharged from the bed .
The adsorber system should be operated at the lowest convenent
temperature because of the dependence of adsorptve capacty on tempera-
ture. Beta decay of the fsson product gases whle passng through the
adsorber system wll ncrease the temperature of the adsorber materal
and reduce the adsorptve capacty . Temperature control s especally
crtcal f the adsorber system uses a combustble adsorber materal, such
as actvated charcoal, wth oxygen as the dluent or sweep gas .
6-3.4 HRE-2 fsson product gas adsorber system . The HRE-2 uses a
fsson product gas adsorber system contanng Columba G actvated
charcoal. Oxygen, contamnated wth the fsson product gases, s removed
-
from the reactor, dred, and passed nto ths system, and the effluent gases
are dspersed nto the atmosphere through a stack .
The adsorber system contans two actvated charcoal-flled unts con-
nected parallel to the gas lne from the reactor . Each unt conssts of
40 ft of z-n . ppe, 40 ft of 1-n . ppe, 40 ft of 2-n . ppe, and 60 ft of 6-n .
ppe connected n seres . The entre system s contaned n a water-flled
pt, whch s bured underground for gamma sheldng purposes . Each
unt s flled wth approxmately 520 11) of Columba G actvated charcoal,
8 to 14 mesh .
The heat due to beta decay of the short-lved krypton and xenon so-
topes s dmnshed by an empty holdup volume composed of 160 ft of
3-n . ppe between the reactor and the charcoal adsorber system . Ths pre-
vents the temperature of the charcoal n the nlet sectons of the adsorber
system from exceedng 100C . Excessve oxdaton of the charcoal by the
oxygen n the gas s further prevented by water-coolng the beds .
Before the adsorber system was placed n servce, ts effcency was
tested under smulated operatng condtons [9] . A pulse of Krss (23
mllcures) was njected nto each unt of the adsorber system and swept
through wth a measured flow of oxygen . In ths way the krypton holdup
tme was determned to be 30 days at an oxygen flow rate of 230 cc/mn/
unt. Based on laboratory data from small adsorber systems, the holdup
tme for xenon s larger than that for krypton by a factor that vares from
30 to 7 over the temperature range of 20 to 100 C. From these data, t s
estmated that the maxmum temperature of the HRE-2 adsorber system
wll vary between 20 and 98C after the reactor has been operatng at
10 Alw power level long enough for the gas composton and charcoal
temperature to have reached equlbrum through the entre length of the
adsorber unt. The holdup performance of the adsorber system was cal-
culated wth correctons for the ncreased temperature expected from the
fsson gases . The calculated holdup tme was found to be 23 days for
krypton and 700 days for xenon ; ths would permt essentally no Xe
133
to escape from the trap .
6-4. CORE PROCESSING : SOLUBLES
6-4
.1 Introducton . Whle the solds-removal scheme dscussed n Sec-
ton 6-1 wll lmt the amount of solds crculatng through the reactor sys-
tem, soluble elements wll buld up n the fuel soluton . Nckel and man-
ganese from the corroson of stanless steel and fsson-produced cesum
wll not precptate from fuel soluton under reactor condtons untl con-
centratons have been reached whch would result n fuel nstablty and
loss of uranum by coprecptaton . Loss of neutrons to these posons
would serously decrease the probablty of the reactor producng more
-
fuel than t consumes. Therefore, a volume of fuel soluton suffcent to
process the core soluton of the reactor at a desred rate for removal of
soluble materals s dscharged along wth the nsoluble materals concen-
trated nto the hydroclone underflow pot. Ths rate of removal of soluble
materals depends on the nature of other chemcal processng beng done
and on the extenWof corroson. For example, operaton of an odne re-
moval plant (Secton 6-5) reduces the buldup of cesum n the fuel to an
nsgnfcant value by removng cesum precursors.
6-4.2 Solvent extracton. Processng of the core soluton of a homo-
geneous reactor by solvent extracton s the only method presently aval-
able whch has been thoroughly proved n practce . However, the amount
of uranum to be processed daly s so small that operaton of a solvent ex-
tracton plant just for core soluton processng would be unduly expensve.
Therefore, the core soluton would normally be combned wth blanket ma-
teral from a thermal breeder reactor and be processed through a Thorex
plant, but wth a plutonum-producng reactor separate processng of core
and blanket materals wll be needed. These process schemes are dscussed
n detal n Sectons 6-6 and 6-7 .
The uranum product from ether process would certanly be satsfactory
for return to the reactor . Snce sold fuel element refabrcaton s not a
problem wth homogeneous reactors, decontamnaton factors of 10 to 100
from varous nucldes are adequate and some smplfcaton of present
solvent extracton schemes may be possble .
6-4.3 Uranyl peroxde precptaton. A process for decontamnatng the
uranum for quck return to a reactor has been proposed as a means of
reducng core processng costs. A conceptual flowsheet of ths process,
whch depends on the nsolublty of U04 under controlled condtons for
the desred separaton from fsson and corroson products, s shown n
Fg
. 6-10. A prerequste for use of ths scheme s that losses due to the
nsoluble uranum contaned n the solds concentrated n the hydroclone
plant be small . However, laboratory data obtaned wth synthetc solds
smulatng those expected from reactor operaton ndcate that the uranum
content of the solds wll be less than 1%o by weght. Verfcaton of the
results wll be sought durng operaton of the HRE-2.
In the proposed method, the hydroclone system s perodcally solated
from the reactor and allowed to cool to 100 C. The hydrolyzed solds re-
man as such, but the rare-earth sulfate solds concentrated n the under-
flow pot redssolve upon coolng. The contents of the underflow pot are
dscharged to a centrfuge where solds are separated from the uranum-
contanng soluton and washed wth D20, the suspenson beng sent to a
waste evaporator for recovery of D20.
-
FIG. 6-10 .
Schematc flow dagram for decontamnatng uranum by uranyl
peroxde precptaton .
Uranum n the clarfed soluton s precptated by the addton of ether
D202 or \a202. By controllng pD and precptaton condtons, a fast
settlng precptate can be obtaned wth less than 0 .1 1/
'
c
of the uranum
remanng n soluton . The U04 precptate s centrfuged or fltered and
washed wth D20 and dssolved n 50% excess of D 2S0 4 at 80 C before
beng returned to the reactor .
In laboratory studes uranum losses have been consstently less than
0
.1% for ths method and decontamnaton factors from rare earths greater
than 10. Decontamnaton factors from nckel and cesum have been 600
and 40, respectvely. It s estmated that the product returned to the re-
actor would contan about 20 ppm of sodum as the only contamnant n-
troduced durng processng . Although ether the addton of D202 or use
of D202 generated by radaton from the soluton tself appears attractve,
acd lberated by the precptaton of U04 must be neutralzed f uranum
losses are to be mnmzed . Snce the entre operaton s done n a D20
system, no specal precautons to avod contamnatng the reactor wth
ordnary water are needed .
6-5 . CORE
PROCESSING : IODINE*
6-5.1 Introducton . The removal of odne from the fuel soluton of a
homogeneous reactor s desrable from the standpont of mnmzng the
bologcal hazard and neutron posonng due to odne and reducng the
producton of gaseous xenon and ts assocated problems . Iodne wll also
*Contrbuton from S . Peterson .
-
poson platnum catalysts [10] used for radolytc gas recombnaton n
the reactor low-pressure system and may catalyze the corroson of metals
by the fuel soluton . For ths reason a consderable effort has been carred
out at OIINI, and by Vtro [11] to nvestgate the behavor of odne n
soluton and to develop methods for ts removal. In ths regard, the odne
sotopes of prmary nterest are 8-day131
and 6 .7-hr 1 13 `'
6-5
.2 The chemstry of odne n aqueous solutons. Iodne n aqueous
soluton at 25 C can exst n several oxdaton states. The stable speces
are odde on, I - ; elemental odne, 12 ;odate, 10,;- ; and perodate, 104
or H510(j. The last of these exsts only under very strongly oxdzng con-
dtons, and s mmedately reduced under the condtons expected for a
homogeneous reactor fuel. Iodde on can be formed from reducton of
other states by metals, such as stanless steel, but n the presence of the
oxygen necessary n a reactor system t s readly converted to elemental
odne ; ths converson s especally rapd above 200 C . Thus the only
states of concern n reactor fuel solutons are elemental odne and odate .
Under the condtons found n a hgh-pressure fuel system the odne s
largely, f not all, n the elemental form .
Volatlty of odne . Snce the volatle elemental state of odne s pre-
domnant under reactor condtons, the volatlty of odne from fuel solu-
ton s the bass for proposed odne-removal processes. The vapor-lqud
dstrbuton coeffcent [11] (rato of mole fracton of odne n vapor to
that n lqud) for smulated fuel soluton and for water at the temperatures
expected for both the hgh-pressure and low-pressure systems of homo-
geneous reactors s gven n Table 6-6 .
TABLE 6-6
VAPOR-LIQUID DISTRIBUTION OF IODINE
Dstrbuton coeffcent,
vapor/lqud
Soluton
Hgh pressure Low pressure
(260-330C) (100C)
Clean fuel soluton 7 .40.34
(0 .02 m U02SO4-0 .005 m H2SO4-
0 .005
M
CuSO4, 1-100 ppm 12
Fuel soluton wth mxed fsson and
corroson products2.4
Water (pH 4 to 8, 1-13 ppm 1 2 )0 .29 0.009
-
FIG. 6-11 . Vtro odne test loop .
A number of conclusons are evdent from these data . Iodne s much
more volatle from fuel soluton than from water at ether temperature .
Fsson and corroson products appear to ncrease the volatlty of odne
from fuel soluton at 100 C. Increasng the temperature from 100 to 200 C
ncreases the volatlty of odne relatve to that of water. No systematc
varaton of odne volatlty has been found wth odne concentraton n
the range 1 to 100 ppm or temperature n the range 260 to 330 C .
The volatlty of odne from smulated fuel soluton has been verfed
by experments n a hgh-pressure loop, shown schematcally n Fg . 6-11
[I]]. The crculatng soluton was contacted wth oxygen n the ejector ;
the separated gas was strpped of odne by passng through a bed of sl-
vered alundum whch was superheated to prevent steam condensaton .
Potassum odde soluton (contanng a radoactve tracer, 1 131 ) was
rapdly njected nto the loop to gve an odne concentraton of 10 ppm .
The odne concentraton decreased exponentally wth tme n the crcu-
latng soluton. Table 6-7 gves the half-tmes for odne removal and the
volatlty dstrbuton coeffcent, calculated from the removal rate and the
flow rates, based on three experments wth clean fuel soluton and two
wth added ron . Wthn the accuracy of flow rate measurement, the coef-
-
fcent agrees wth the average value of 7.4 obtaned n numerous statc
tests over the hgh-temperature range . Iron appears to have no effect .
Oxdaton state of odne at hgh temperatures and pressures . Whle odate
on s qute stable at room temperature, at elevated temperatures t de-
composes accordng to the equlbrum reacton
4103
--}-411
+ {
212+502+21120 .
The extent of ths decomposton n uranyl-sulfate solutons above 200 C
s not known wth certanty, snce all observatons have been made on
samples that have been wthdrawn from the system, cooled, and reduced
n pressure before analyss . Although the odne n such samples s prn-
cpally elemental, some odate s always present, possbly because of re-
versal of the odate decomposton as the temperature drops n the sample
lne. Such measurements therefore gve an upper lmt to the odate con-
tent of the soluton. If perodate s ntroduced nto uranyl-sulfate soluton-
at elevated temperatures, t s reduced before a sample can be taken to
detect ts presence. Iodde smlarly dsappears f an overpressure of oxy-
gen s present, although odde to the extent of 401-/, of the total odne has
been found n the absence of added oxygen [11] .
Methods that have been used for determnng the odne/odate rato n
fuel solutons are (a) analyss of samples taken from an autoclave at 250 C
at measured ntervals after njecton of odne n varous states [11],
(b) analyss of samples taken from the lqud n lqud-vapor equlbrum
studes at 260 to 330C [11], (c) rapd samplng from statc bombs at
250 to 300C [12], and (d) contnuous njecton of odate-contanng fuel
soluton nto the above descrbed ejector loop at 220 C and determnng
TABLE 6-7
IODINE REMOVAL FROM A HIGH-PRESSURE Loop
Soluton
Tempera-
turn,
C
Iodne
removal
half-tune,
mn vapor/
Iodne
dstrbuton
coeffcent,
lqud
Clean fuel soluton 230 13 .0 7 .6
(0 .02 1nU0
2S0 .1-0 .005 n H2SO4-
6 .5 16 .8
0 .005 M CUSO4) 13 .0
8 .4
Fuel + 30 ppm Fee+ 220 11 .0 10 .9
Fuel + 300 ppm Fe -3+ 225 11 .0 9 .5
-
oxdaton states n samples wthdrawn [11] . The odne/odate rato n
these samples has vared from slghtly over 1 to about 70, wth no apparent
relaton to varatons n temperature, oxygen pressure, and total odne
concentraton .
The strongest ndcaton of odate nstablty was n the loop exper-
ments, whch gave the hghest observed odne/odate rato, even though
odne was contnuously ntroduced nto the flowng stream as odate
and removed by oxygen scrubbng as elemental odne . The low odate
content of the samples from these experments corresponded to a frst-
order odate decomposton rate constant of 6 .2 mn' . Iodate con-
tents averagng about 10% of the total odne have been observed n
0.04 m U02S04-0.005 m CuSO4-H2SO4 soluton, rapdly sampled from
a statc bomb through an ce-cooled ttanum sample lne . The observed
odate content was unrelated to whether the free sulfurc acd concentra-
ton was 0.02 or 0 .03 m, whether the temperature was 250 or 300 C, and
whether or not the soluton was exposed to cobalt gamma radaton at an
ntensty of 1 .7 watts/kg .
Oxdaton state of odne at low temperatures . At 100C the odate de-
composton and odne oxdaton are too slow for equlbrum to be es-
tablshed n reasonable perods of tme . Thus both states can persst under
smlar condtons . In stanless-steel equpment both states are reduced to
odde, whch s oxdzed to odne f oxygen or odate s present [12] .
In a radaton feld the odde s oxdzed, odne s oxdzed f suffcent
oxygen s present, and odate s reduced [13] . At the start of rradaton,
odate s reduced, but n the presence of suffcent oxygen, odne s later
reoxdzed to odate, probably by radaton-produced hydrogen peroxde
whch accumulates n the soluton . Fnally, a steady state s reached wth a
proporton of odate to total odne whch s ndependent of total odne con-
centraton from 10-6 to 10 m and temperatures from 100 to 110 C, but
strongly dependent on uranum and acd concentratons and on the hydro-
gen/oxygen rato n the gas phase . When the temperature s ncreased to
120C there s a marked decrease n odate stablty under all condtons of
gas and soluton composton. Expermental data on the effects of radaton
ntensty, temperature, and gas composton for the rradaton of a typcal
fuel soluton contanng 0 .04 m U 02SO4-0.01 m H2SO4-0.005 M CuSO4
are gven n Ref . 13. The steady-state odate percentages are also gven n
ths reference .
6-5
.3 Removal of odne from aqueous homogeneous reactors . It s
clear that under the operatng condtons of a power reactor, odne n the
the fuel soluton s manly n the volatle elemental state . It can therefore
be removed by sweepng t from the soluton nto a gas phase, strppng
-
t from the gas stream by trappng t n a sold absorber or by contactng the
gas wth a lqud .
Numerous experments have shown that slver supported on alundum s
a very effectve reagent for removng odne from gas or vapor systems,
although ts effcency s consderably reduced at temperatures below
150C . Slver-plated Yorkmesh packng s very effectve for removng
odne from vapor streams n the range 100 to 120C. In one n-ple ex-
perment [14] 90% of the fsson-product odne was concentrated n a
slvered-alundum pellet suspended n the vapor above a uranyl-sulfate
soluton . Ths method of usng a sold odne absorber, however, would
present dffcult engneerng problems, snce xenon resultng from odne
decay would be expected to leave the absorber and return to the core unless
the absorbers were solated after short perods of use and remotely replaced.
Iodne removal by gas strppng requres a contnuous fuel letdown . In
case ths s not desrable, the vapor can be strpped of odne n the hgh-
pressure system by contactng wth a small volume of lqud whch s sub-
sequently dscharged . Lquds consdered nclude water and aqueous
solutons of alkal, sodum sulfte, or slver sulfate [1.1] . Although the so-
lutons are much more effectve odne strppers than pure water, ther use
requres elaborate provson for preventng entranment n the gas and sub-
sequent contamnaton of the fuel soluton . Thus most of the effort n
desgn of odne-removal systems s based on strppng by pure heavy
water .
One possble odne-removal scheme uses 0 2 or02 + D2
strppng [15] .
The odne s scrubbed from the fuel soluton by the gas n one contactor
and then strpped from the gas by heavy water n a second contactor . Ths
water would then be let down to low pressure and stored for decay or proc-
essed to remove odne .
In most homogeneous reactors some of the fuel soluton s evaporated
to provde condensate for purge of the crculatng pump and pressurzer .
Snce odne s strpped from the fuel by ths evaporaton ths operaton can
be used for odne removal . Ths method, whch s llustrated n Fg . 6-12,
has been proposed for the HRE-3 [16] . Here a stream of the fuel soluton
s scrubbed wth oxygen n the pressurzer. The steam s condensed and the
oxygen recycled . The condensate s dstlled to concentrate the odne nto
such a small volume that ts letdown does not complcate reactor operaton .
Iodne removal n the HRE-2 . Iodne adsorpton on the platnzed alu-
mna recombnaton catalyst, such as that used n the HRE-2, posons the
catalyst severely [10] . Although the catalyst can be restored by operaton
at 650C, ths would not be feasble n HRE-2 operaton . A method for
removng odne from the gas stream by contact wth alundum or York-
mesh coated wth slver was developed n the HRT mockup . Iodne was
ntroduced nto the system and vapor from the letdown stream and dump
-
FIG. 6-12 . Iodne removal system proposed for HRE-3 .
tank was passed through a slvered alundum bed and the recombner, and
then to a condenser . Condensate was returned to the hgh-pressure loop
through a pressurzer and the crculatng pump . After njecton, the odne
concentraton of the hgh-pressure loop dropped from 1.8 mg/lter to
0.1 mg ,'lter n 2 hr . In smlar experments wth slvered Yorkmesh, odne
levels n the condensate and pressurzer were even lower relatve to the
hgh-pressure loop . The Yorkmesh effcency depended strongly on how
densely t was packed . The odne removal effcences calculated from
these experments and others are gven n Table 6-8 . In laboratory ex-
perments wth a 1-n .-dameter bed whch could not be tghtly packed,
Yorkmesh effcences were consstently poorer than those of slvered
alundum .
The ablty of a bed of slver-plated Yorkmesh to remove odne from the
reactor system was apparently confrmed durng the ntal operatng
perod of the HRE-2 [17] . Here the odne actvty n the reactor fuel
appeared to be even lower than expected when odne was removed at
the same fractonal rate as fuel soluton was let down from the hgh-
pressure system . Less than 3% of the odne produced durng 40 Mwh of
operaton was found n the fuel soluton . Experence wth the HRT
mockup ndcates that the odne not n soluton was held on the slvered
bed .
-
6-6 . URANYL SULFATE BLANKET PROCESSING*
6-6.1 Introducton.
The uranyl sulfate blanket soluton of a plutonum
producer s processed to remove plutonum and to control the neutron
posonng by corroson and fsson products. Although a modfed Purex
solvent extracton process can be used for plutonum removal, the method
shown schematcally n Fg. 6-3, based on the low solublty of plutonum
n uranyl sulfate soluton at 250 C, appears more attractve. A hydroclone
smlar to that used for reactor core processng s used to produce a con-
centrated suspenson of Pu02 along wth sold corroson and fsson prod-
ucts . The small volume of blanket soluton carryng the plutonum s
evaporated to recover the heavy water and the solds are dssolved n
ntrc acd. After storage to allow Np 23s to decay, plutonum s decon-
tamnated by solvent extracton .
6-6.2 Plutonum chemstry n uranyl sulfate soluton . The amount of
plutonum remanng dssolved n 1 .4 m U02S04 at 250 C s dependent
on a number of varables, ncludng soluton acdty, plutonum valence,
and ntal plutonum concentraton . Under properly controlled cond-
tons, less than 3 mg/kg H2O has been obtaned . Snce plutonum s re-
moved from soluton by hydrolyss to Pu02, solubltes are ncreased by
ncreasng the acdty . Table 6-9 summarzes data on the solublty be-
havor of plutonum for varous acdtes.
*Contrbuton from R . E. Leuze .
TABLE 6-8
IODINE REMOVAL EFFICIENCY OF SILVERED BEDS IN HRT MOCKUP
Absorber
Bed heght,
n.
Temperature,
C
Effcency,
%
Slvered alundum 8150 97 .7
rngs 8 12081 .0
5 110
64 .0
Yorkmesh, 22 lb/ft 3 10 12097 .0
Yorkmesh, 29 lb/ft 3 6120 99 .6
-
Plutonum behavor s dffcult to predct because of ts complex valence
pattern . In the absence of rradaton, plutonum dssolved n 1 .4 m U02SO4
under a stochometrc mxture of hydrogen and oxygen at 250 C exsts n
the tetrapostve state . However, when dssolved chromum s present or
when an overpressure of pure oxygen s used, part of the plutonum s ox-
dzed to the hexapostve state. Experments ndcate that n the presence
of Co60 gamma rradaton [18], reducng condtons preval even under an
oxygen pressure and plutonum s held n the tetrapostve state . The
valence behavor dscussed here s somewhat n queston, snce actual
valence measurements were made at room temperature mmedately after
coolng from 250C. It s known that tetrapostve plutonum wll dspro-
portonate upon heatng [19] . The dsproportonaton n a sulfate system
s depressed by the sulfate complex formaton wth tetrapostve plu-
tonum . These results ndcate that plutonum n a reactor wll be pre-
domnantly n the tetrapostve state .
When the plutonum concentraton exceeds the solublty lmt, plu-
tonum wll hydrolyze to form small partcles of Pu02 about 0 .5 mcron
n dameter and n pyrex, quartz, or gold equpment forms a loose prec-
ptate wth neglgble amounts adsorbed on the walls . However, f these
solutons are contaned n type-347 stanless steel, ttanum, or Zrcaloy,
a large fracton of the Pu02 adsorbs on and becomes ncorporated wthn
the oxde corroson flm . Attempts to saturate these metal surfaces wth
plutonum n small-scale laboratory experments were unsuccessful even
though plutonum adsorpton was as much as 1 mg/cm 2 .
6-6.3 Neptunum chemstry n uranyl sulfate soluton . Neptunum ds-
solved n 1.4 m U02SO4 at 250 C under ar, stochometrc mxture hy-
drogen and oxygen, or oxygen s stable n an oxdzed valence state, prob-
TABLE 6-9
SOLUBILITY OF TETRAVALENT
PLUTONIUM
IN 1
.4 m U02SO4 AT 250C
Excess sulfurc acd,
m
Pu(IV) solublty,
mg/kg H20
0 3 .7
0 .1 17
0 .2 39
0 .3 68
0 .4 105
-
ably Np(V). The solublty s not known, but t s greater than200 mg/kg
H20. Snce the equlbrum concentraton s only about 50 mg/kg H2
0,
for a 1 .4 m U02SO4 blanket wth an average flux of 1.8 X 10 14 neu-
trons/(cm 2)(sec),all the neptunum should reman n soluton n most
reactor desgns .
6-6.4 Plutonum behavor under smulated reactorcondtons. Pluto-
num behavor n actual uranyl sulfate blanket systems has not been
studed ; however, small-scale statc experments wth 100 ml of soluton
and crculatng loop experments wth 12 lters of soluton have been car-
red out n the absence of rradaton under condtons smlar to those ex-
pected n an actual reactor .
In the statc experments, plutonum was added batchwse to 1 .4 m
U02SO4 at a rate of about 6 mg/kg H20/day . The soluton was heated
overnght n a pyrex-lned autoclave at 250C under 200 ps hydrogen and
100 ps oxygen. The soluton was cooled to room temperature for analyss
and for addng more plutonum . Ths was repeated untl a total of 140 mg
of plutonum per klogram of water was added . Small dsks of type-347
stanless steel were suspended n the soluton throughout the experment to
determne the amount of plutonum adsorpton . The behavor of plu-
tonum for a stanless-steel surface area/soluton volume rato of 0.6 cm 2/ml
s shown n Fg . 6-13. As the plutonum concentraton was gradually n-
creased to 45 mg/kg H2O, essentally all the plutonum remaned n solu-
ton as Pu(VI) . There was a small amount of adsorpton, but no precpta-
ton. Durng the next few addtons the amount of plutonum n soluton
decreased rapdly to about 5 mg/kg H2O. At the same tme there was a
rapd ncrease n plutonum adsorpton and n the formaton of a loose
Pu02 precptate . All plutonum added after ths was ether adsorbed or
precptated .
Other experments were made wth surface/volume ratos of0.2 and
0.4 cm 2/ml. In all cases, the plutonum remanng n soluton and the plu-
tonum adsorpton per square centmeter were essentally the same as
that shown n Fg. 6-13. Thus, by decreasng the surface/volume rato,
t s possble to ncrease the amount of plutonum n the loose precptate .
For example, when the total plutonum addton was 130 mg/kg H2O, 40%
of the plutonum was as a loose precptate for a surface/volume rato of
0.6 cm 2/ml, 60% for a rato of 0.4 cm 2/ml, and 68% for a rato of
0.2 cm 2/ml .
Plutonum behavor under dynamc condtons was studed by njectng
dssolved plutonum sulfate and preformed PuO2 nto a crculatng stream
of 12 lters of 1 .4 m U02SO4 at 250C under 350 ps oxygen. Ths soluton
was contaned n a type-347 stanless steel loop equpped wth a canned
rotor pump, a hydroclone, metal adsorpton coupon holders, and a small
-
FIG. 6-13 . Plutonum behavor n uranyl sulfate soluton contaned n type-347
stanless steel .
pressure vessel that could ,be connected and removed whle the loop was n
operaton. Plutonum was added and crculatng-soluton samples were
taken through ths vessel . Tetrapostve plutonum added to the crculatng
soluton was completely oxdzed to hexapostve n less than 5 mn. When
45 mg/kg H2O of dssolved plutonum was added every 8 hr, the amount
of plutonum crculatng n soluton ncreased to a maxmum of about
150 mg/kg 1120 . As more plutonum was added, t was rapdly adsorbed
on the loop walls . After the last addton of plutonum, the loop was
operated at 250C for several days. Twelve hours after the last addton
the plutonum concentraton had decreased to 100 mg/kg H2O, and about
40 hr later the amount of plutonum n soluton had dropped to an ap-
parent equlbrum value of 60 mg/kg1120
. Essentally all the plutonum
removed from soluton was adsorbed on equpment walls unformly
throughout the loop. Less than 0 .1% of the plutonum was removed n
-
the hydroclone underflow, and no precptated solds were crculatng .
Even when 850 mg of plutonum as preformed Pu02 was njected nto the
loop, no crculatng solds were detected 5 mn later . Only 20% of ths plu-
tonum was removed by the hydroclone, 35% was adsorbed on the stanless
steel, and the rest was dstrbuted throughout the horzontal sectons of
the loop as loose solds . The hydroclone was effectve for removng solds
that reached t, but the loop walls and low velocty n horzontal ppes
were effectve traps for Pu02 .
There are several dfferences n condtons between the loop runs and an
actual reactor, the most mportant of whch are probably the presence of
radaton, the lower surface/volume rato (0 .4 compared wth 0 .8 cm2/ml
for the loop), the slower rate of plutonum growth n the reactor (12 to
15 mg/kg H2O/day) and the probablty that a plutonum producer
would have to be constructed of ttanum and Zrcaloy to contan the con-
centrated uranyl-sulfate soluton . Based on these laboratory results, how-
ever, t appears that plutonum adsorpton on metal walls may be a serous
obstacle to processng for removal of precptated Pu02 .
6-6.5 Alternate process methods . Because of the problem of plutonum
adsorpton on metal walls, removal methods based on plutonum concen-
tratons well below the solublty lmt have been consdered . In a full-
scale reactor plutonum wll be formed at the rate of up to 12 to 15
mg/kg H2O/day . In order to keep the plutonum concentraton below
3 mg/kg H20, the entre blanket soluton must be processed at least four
to fve tmes a day . By addng 0.4 m excess H2SO4 (see Table 6-9), the
plutonum solublty s ncreased to greater than 100 mg/kg H2O and the
blanket processng rate can be decreased to once every 3 or 4 days . Slghtly
longer processng cycles can be used f part of the plutonum s removed as
neptunum before t decays .
Of the varous alternate processes consdered, on exchange and ad-
sorpton methods show the most promse . Dowex-50 resn, a strongly
acdc sulfonc acd resn loaded wth U02++ ,
completely removed tetra-
postve plutonum from 1 .4 m U02SO4 contanng 20 mg of plutonum
per lter [20] . The resn capacty under these condtons, however, has not
been determned . Because of the hgh radaton level t may not be feasble
to use organc resns. Sorpton of plutonum on norganc materals shows
some possbltes as a processng method [21] . Although rather low plu-
tonum/adsorber ratos have been obtaned, ndcatons are that capactes
wll be sgnfcantly hgher at hgher plutonum concentratons . Specal
preparaton of the adsorbers should also ncrease capactes . Attempts to
coprecptate plutonum wth tr- or tetrapostve odates, sulfates, oxalates,
and arsenates were not successful, owng to the hgh solubltes of these
materals n 1 .4 m U02SO4 .
-
FIG
. 6-14
.
Thorex process, feed preparaton flowsheet .
FIG. 6-15 . Thorex process, solvent extracton co-decontamnaton flowsheet .
-
FIG. 6-16 .
Thorex process, uranum solaton and thrd cycle flowsheet .
6-7
. THORIUM OXIDE BLANKET PROCESSING
6-7
.1 Introducton . At the present the only practcal method avalable
for processng rradated thorum-oxde slurry s to convert the oxde to a
natural water-thorum ntrate soluton and treat by the Thorex process.
Although ths method s adequate, t s expensve unless one plant can be
bult to process thorum oxde from several full-scale power reactors .
Therefore methods for Th02 reprocessng whch could be economcally n-
corporated nto the desgn and operaton of a sngle power staton have
been consdered. Alternate methods that have been subjected to only bref
scoutng-type expermentaton are dscussed n Artcle 6-7.3 .
6-7.2 Thorex process.* The Thorex process has been developed to sep-
arate thorum, U 233 ,fsson product actvtes, and
Pa233 ;
to recover the
thorum and uranum as aqueous products sutable for further drect
handlng; and to recover sotopcally pure
U233
after decay-storage of the
Pa
233
The flowsheet ncludes two solvent-extracton cycles for thorum
and three solvent-extracton cycles plus on exchange for the uranum.
Although only rradated thorum metal has been processed, the process s
expected to be satsfactory for recovery of thorum and uranum from
homogeneous reactor fuels.
The Thorex process may he dvded nto three parts : feed preparaton,
*Contrbuton from '. T. McDuffee .
-
solvent extracton, and productconcentraton and purfcaton . These
three dvsons are shown n Fgs .6-14, 6-15, and 6-16 .
In the feed preparaton step, uranyl sulfate soluton from the reactor core
and thorum oxde from the blanket system, freed of D20 and suspended n
ordnary water, are fed nto the dssolver tank . The dssolvent s 13 N
ntrc acd to whch has been added catalytc amounts (0 .04 N) of sodum
fluorde. When short-cooled thorum s beng processed, potassum odde
s added contnuously to the dssolver to provde for sotopc dluton of
the large amount of fsson-produced 11 31 whch s present. The dssolver
soluton s contnuously sparged wth ar, and the volatlzed odne s re-
moved from the off-gases n a caustc scrubber .
The dssolver soluton s transferred to the feed adjustment tank where
alumnum ntrate s added, excess ntrc acd recovered, and the resultant
soluton made slghtly acd-defcent by evaporatng untl a temperature
of 155C s reached. Durng dgeston n the feed adjustment tank any
slca present s converted to a form that wll not cause emulson problems
n pulse columns, and fsson products generally are converted to forms
less lkely to be extracted by the solvent (42%0 TBP n Amsco) .
In the solvent extracton step thorum and uranum are co-extracted n
the frst cycle; subsequent parttonng of thorum and uranum n the
second cycle gves two decontamnatng cycles to both products whle
usng only fve columns . For short-decayed thorum a reductant, sodum
hydrogen sulfte, s contnuously added to the feed streams of both cycles
to decrease the effect of ntrte formed by rradaton. Wthout the sulfte
addton, the ntrte formed by radaton decomposton of ntrates con-
verts ruthenum to a solvent-extractable form . Acd defcency n the
second cycle feed s acheved by addng dbasc alumnum ntrate (dban) .
The spent organc from the second cycle s recycled to the frst cycle as
the organc extractant . The spent solvent from the frst cycle s processed
through a solvent-recovery system and reused as the organc extractant n
the second cycle .
In the uranum product concentraton and purfcaton step (Fg . 6-16),
uranum s solated by on exchange, usng upflow sorpton and downflow
eluton. In ths way a concentrated uranum soluton n 6 N HN03 s ob-
taned. Ths soluton s stable enough for storage or s sutable as a feed
for the thrd uranum extracton cycle . The thrd uranum cycle s a
standard extracton-strppng solvent-extracton system usng 15% TBP-
Amsco as the organc extractant. Although nstalled as a part of the com-
plete Thorex flowsheet, the thrd cycle may be used separately for re-
processng long-stored uranum to free t of objectonable decay daughters
of U232 . When used as an ntegral part of the Thorex scheme, addtonal
decontamnaton of the uranum s acheved and the ntrate product s
well adapted for extended storage or future reprocessng .
-
TABLE
6-10
AVERAGE DECONTAMINATION FACTORS
FOR THORIUM AND URANIUM
PRODUCTS
IN THE
THOREX PILOT PLANT
Thorum rradatedto 3500
one addtonal
grams of mass-233
uranum
per
cycle
ton, two complete
for materal
cycles
decayed only
for both uranum and
30 days .
thorum,
Decontamnaton factors
GrossPa Ru Zr-A b
Total rare earths I
Thorum
400 days decayed 1 x 105
1 x 10 4 4 x 103 3 x 10 5 2 x 106
30 days decayed 4 x 10 47 x 10 6 200 3 x 10 4 2 x 106
9 x 108
Uranum-233
400 days decayed 3 x 103 x 10
5
2 x 105 8 x 10 .5 9 x 10 8
30 days decayed 5 x 10'5 x 10 10 4 x 100 7 x 10 6 3 x 10 8
3 x 10 7
-
For return to an aqueous homogeneous reactor the decontamnated
uranum would probably be precptated as the peroxde, washed free of
ntrate, and then dssolved n D2SO4 and D20 . Product thorum would
be converted to thorum oxde by methods descrbed n Secton 4-3 .
The adaptablty of the Thorex flowsheet just descrbed to processng
thorum rradated to contan larger amounts of
U233
per ton and decayed
a short tme has been demonstrated n the Thorex Plot Plant at Oak Rdge
Natonal Laboratory [22]. Ffteen hundred pounds of thorum rradated
to 3500 grams of
U233
per ton and decayed 30 days was processed through
two thorum cycles and three uranum cycles . The decontamnaton fac-
tors for varous elements acheved wth short-decayed materal are com-
pared n Table 6-10 wth results obtaned wth longer-decayed materal .
Whle the decontamnaton factors obtaned wth the short-decayed ma-
teral compare favorably wth the factors for the long-decayed materal,
the ntal actvty n the short-decayed thorum was 1000 tmes greater
than n the long-decayed . Therefore, whle the thorum and uranum
products dd not meet tentatve specfcatons after two complete cycles,
the uranum product dd meet those specfcatons after the thrd uranum
cycle. Snce the chemcal operatons necessary to convert these materals
to forms sutable for use n a homogeneous reactor can be carred out re-
motely, the products are satsfactory for return to a homogeneous reactor
after two cycles .
6-7.3 Alternate processng method .* Attempts to leach protactnum
and uranum produced n Th02 partcles by neutron rradaton [23] n-
dcate that both are rather unformly dstrbuted throughout the mass of
the Th02 partcle, and mgraton of such ons at temperatures up to 300 C
s extremely slow . Snce calculatons show that the recol energy of frag-
ments from 15
233
fsson s suffcently large to eject most of them from a
partcle of Th02 not larger than 10 mcrons n dameter, ths offers the
possblty of separatng fsson and corroson products from a slurry of
Th02 wthout destroyng the oxde partcles . Such a separaton, however,
depends on the ablty to remove the elements that are subsequently ad-
sorbed on the surface of the Th02 . Adsorpton of varous catons on Th02
and methods for ther removal are dscussed n the followng paragraphs .
Trace quanttes of such nucldes as Zr 95 Nd
147
Y91 , and Ru
103
when
added to a slurry of Th02 n water at 250 C are rapdly adsorbed on the
oxde partcles, leavng less than 10
-4
% of the nucldes n soluton . The
tracer thus adsorbed cannot be eluted wth hot dlute ntrc or sulfurc acd .
The adsorpton of macroscopc amounts of uranum or neodymum on
Th02 at 250C s less for oxde fred to 1600 C than for 650C-fred oxde,
*Contrbuton from R . E. Leuze .
-
TABLE 6-11
EFFECT OF CALCINATION TEMPERATURE ON
URANIUM AND NEODYMIUM ADSORPTION ON TH02 AT 250C
0
.5 g of Th02 slurred at 250C n 10 ml of 0 .005 m Nd(N03)3
or 0.05 m U02SO4-0 .05 m H2SO4 .
Calcnaton temperature,
C
Adsorpton, mg/g
U
Th *
Nd
650 3 .3-4
.4 7 .4
850
1
.9-2
.4
6 .1
1000 0 .72-1 .10
2.4
1100 0 .08-0 .19 0 .5
1600 0 .06-0 .12 0.3
*Sngle numbers represent data from sngle experments . In other cases the range
for several experments s gven .
TABLE 6-12
USE OF PB0 TO DECREASE CATION ADSORPTION ON TH02
0.2 g of Th02 plus varous amounts of PbO coslurred n 10 ml of soluton at
250C for 8 hr .
Solds
PbO/Th02
wt. rato
Soluton
Caton adsorbed on Th02,
ppm
Th02 0
.002 m U02SO4 3100
PbO + Th02 0 .2 0 .002 m U02SO4220
Th02 0 .001 m Ce(N03)36200
PbO + Th0
2
0 .2 0 .001 7n Ce(- 03)3 10
Th0
2
0 .01 n Nd tartrate 2700
PbO + Th02 0 .4 0 .01 m Nd tartrate 10
-
as llustrated n Table 6-11 .Ths change n amount of adsorpton may be
almost entrely due to decrease n surface area of Th02 wth ncreased
frng temperature. The surface area of 1600 C-fred Th02 s only 1 m 2/g
Th02, whle the650C-fred Th02 has a surface area of 35 m 2/g Th02 .
The caton adsorpton on Th02 can be decreased by coslurryng some
other oxde wth the Th02. The added oxde must adsorb fsson products
much more strongly than Th02 and be easly separable from Th02 . The
effectveness of PbO n decreasng caton adsorpton on Th02 s shown n
Table6-12 . When Pb02 was used, more than 99% of the catons added
to the Th02-PbO slurry was adsorbed on the Pb02.
However, catons
adsorbed on Th02 were not transferred to Pb02 when t was added to
slurry n whch the catons were already adsorbed on the Th02 partcles.
Addton of dlute ntrc acd to the Th02-PbO coslurry completely ds-
solved the PbO and the catons adsorbed on t wthout dsturbng the Th02.
In all cases, catons adsorbed on Th02 at250C are so tghtly held that
dlute ntrc or sulfurc acd, even at bolng temperature, wll not remove
the adsorbed materal. However, the adsorbed ons can be desorbed by
refluxng the Th02 n sutable reagents under such condtons that only a
small amount of 1600C-fred Th02 s dssolved . Under the same treat-
ment Th02 fred to only 650C would be 90% dssolved .
-
REFERENCES
1 . A . T
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P. A . HAAS,
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-
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