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

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

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

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 .

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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2 . E . D . ARNOLD andA. T. GRESKY, Relatve Bologcal Hazards of Radatons

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5 .

P. A . HAAS,

Hydraulc Cyclones for Applcaton to Homogeneous Reactor

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Homogeneous

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