I&EC Lecture Series
Chemical Engineering Aspects of Nuclear First in what is planned to be a series of afternoon lectures at national meetings of the ACS each fall, i s this discussion of the challenge to the chemical engineer which lies in the economical development of power
'
from nuclear energy.
In these lectures the Division of Industrial and
. Engineering Chemistry of the American Chemical Society expects to
pmsent s & i s
of timely and unusual interest to a wide segment of
the chemical field.
MANSON BENEDICT Ma&udl*
Idibh of T d d o g y , Cambridpe, Man.
linea of engineering endeavor which must be
If power from nuclear fission is to compete Wnomically with power from coal and oil are:
1. Development of durable and relatively inexpensive rea* tors for camymg out co~trollednuclear fission and utilizing the energy of fieion for production of power. 2. Development of cheap and &cient processes for concenup and Tenuclear fuels and for trsting and using spent uel rom nuclear reactors.
p"%"
The P h c i @ t Y P S of nuclear reactom now under consider* tion for nuclear power planta have been described recently ii declassified versions of reports to the Atomic Energy Commission writken by four teams of industrial companies (3). Much leas has been said about the serond topic, the preparation Of for reactors and the treatment Of materials them. which have This Paper is Concerned with the Principal fuel been for nuclear power PIanta and the chenucal pesw that are needed to make these fuel cycles feaaible and eccnomic. The pmceaees discussed come under the headings of: 1. Isotope separation 2 Extraction of uranium and thorium from their or$a 3. Puriication of uranium and thorium 4. Separation of reactor producta
Finally, a few examples are given of offsbmta from this nuclear chemical technology which may find useful application elsewhere in chemical industry. TI& -on b necessarily somewhat incomplete, partly bec a w pertinent information bas not yet been declassified, but
2x2
more because of actual gaps in our knowledge and experience. These offer chemista and chemical engineers a unique opportunity to help make nuclear power practical.
SU.1 Cycler for Nuclear Power Plants ~~a~~puels. The three fi&nabk isotopes which my be w d as fuel in nuclear resotors are uranium-235, plutonium-239, and uranium-233. Uranium-235 occurs in nature (mixed with 140 parts of uranium-238), whereae plutonium-239 and uranium233 are man-made isotopes, produced by the absorption of neutrona in uranium-238 and thorium-232, respectively. Same of the nuclear properties of these isotopes important in nuclear reactors are given in Table I. The number of neutrons per absorbed deh-as whether nt,clear breeding is paaeble, If tbnumber is greater than 2, it is possible to deab reactor to a net excess of Ikionable material, above that required to keep the reactor in This is as mhen a hionable atom absorbs a neutron and produces two neutrons, one of these is used to up for the ueubon absorbed, and the may be wed to produce a second atom of bsionable material to make up for the atom In thermal reactors, we see that breeding is theoretically passible with uranium-235 and uranium-233, but cannot be carried out with plutonium. In fast reactors, breeding is theoretically possible with plutonium as well 88 uranium-235. The first praotical demonstration of this for uranium-235 hae been given by the Argonne National Lahoratory in the experimental breeder reactor. A fast reactor is one in which the average speed of neutrons is near that which neutrons have at the moment of fission-around 30,000,000 miles per hour. A t these high speeds the probability
INDUSTRIAL AND ENGINEERING CHEMISTRY
Vol. 45, No. 11
c
of a neutron’s being absorbed by a fissionable atom is low, and the neutron absorption cross section, which is a measure of this probability, is small. A thermal reactor is one in which the neutrons have been slowed down until they are in thermal equilibrium with reactor materials, and have speeds around 5000 miles per hour. At these lower speeds, the neutron absorption cross sections are much larger than for fast neutrons, plutonium having a higher cross section than uranium-235. The greater the cross section, the smaller the charge of nuclear fuel required to maintain the fission process. This amount, the critical mass, is lowest for plutonium in thermal reactors, larger for uranium-235 in thermal reactors, and greatest of all in fast reactors. In fast reactors, in particular, it is necessary to use very high concentrations of nuclear fuel, and the amount of dilution of fuel that can be tolerated is severely limited. These nuclear properties of fuels have a determining effect on the fuel cycles which are feasible in nuclear reactors.
Table I.
Fuels for Nuclear Reactors U-235 Pu-239 PIoduced from Occurs in U-238 nature Neutrons produced per neutron absorbed 2 11 1 94 Thermal Fast >2 >2 Keutron absorpbion cross section, barns Thermal 650 1025 Small Small Fast
November 1953
U-233 Thorium
>2
Once-through Uranium Fuel Cycle. The type of power-producing nuclear reactor which received most prominent mention in the reports recently released by the Atomic Energy Commission is a thermal reactor fueled with normal uranium or uranium slightly enriched in uranium-235. These reports show that if neutrons axe slowed down-moderated-with heavy water, natural uranium containing 0.72% uranium-235 may be used as fuel, whereas if graphite is used as moderator, it is advantageous t o use uranium-238 containing about 1% uranium-235 as fuel. To reduce the costs of fuel in such a reactor and minimize chemical processing, W. H. Zinn of the Argonne National Laboratory has proposed a fuel cycle in which natural or slightly enriched uranium is left in the reactor as long as it can sustain the nuclear fission chain reaction, and then is removed and discarded. Let us use such onclassified information as is available to see what fraction of the uranium fuel charge to the reactor could theoretically be burned up in such a cycle. For this purpose we will assume that a thermal reactor is fed with uranium containing 1% uranium-235, and that when operating on fresh fuel the reactor is just critical and produces one atom of plutonium per atom of uranium235 destroyed. Then, of the 2.11 neutrons produced when one neutron is absorbed by uranium-235 in such a reactor, one neutron is absorbed by uranium-238, and 0.11 neutron is lost, either through capture by other elements or leakage from the reactor. The left scale of Figure 1 illustrates the changes in concentration of uranium-235 and plutonium-239 which takes place its this fuel is burned in such a thermal reactor. The extent of burnup is represented by the bottom scale in terms of the percentage of fuel converted to fission products. The lower, right scale shows the amount of heat in megawatt-days produced per killogram of fuel. As the heat of fission of uranium-235 and plutonium-239 is close to 0.9 Mw.-day per gram, 1% conversion of fuel to fission products corresponds to 9 Mw.-days of heat from 1 kg. of fuel. As irradiation progresses, uranium-235 is consumed and its concentration decreases exponentially, since there is no reaction for regenerating it. A t the same time, plutonium is formed b y absorption of neutrons in uranium-238. Its concentration increases, a t first as rapidly as that of uranium-235 decreases, and later mort- slowly, as plutonium begins to undergo fission also. The plutonium concentration finally levels off at about 0.6% when it is being consumed as rapidly as it is being produced. The net effect of these changes in concentration on nuclear reactivity is shown by the top line of the chart. By “excess reactivity” is meant the ratio of the difference between production and consumption of neutrons by a fuel element to the consumption of neutrons by fissionable material in the fuel element when first charged to the reactor. The average excess reactivity of the entire reactor charge must be greater than 1 for the fission reaction to continue-that is, for combustion of the nuclear fuel to be maintained. At f i s t , the excess reactivity increases, owing to the build-up of plutonium with its higher cross.section. As the plutonium concentration levels off, the reactivity begins to decrease because of the continuing build-up of fission products, which absorb neutrons. At 0.6% fuel converted the excem reactivity has dropped to zero and a t 0.97% the average excess reactivity has dropped to zero (area B equals area A ) . I n a batch reactor, in which all the fuel is irradiated uniformly and has the same reactivity, it will be impossible to irradiate the fuel beyond the point a t which the reactivity drops to zero. I n this case, it would be impossible to convert more than 0.6y0of the fuel to fission products before the fuel would no longer sustain the nuclear reaction. On the other hand, in a steady-flow reactor in which fuel moves progressively through the reactor, in principle like coal in a chain-grate stoker, it is theoretically possible to extend the life of the fuel till 0.97% has been converted to fission products. In this case, fuel with all degrees of irradiation between fresh fuel and spent fuel is present in the reactor at the same time, the initial excess reactivity offsets the subsequent reactivity deficiency, and the fuel may be left in the reactor till
INDUSTRIAL AND ENGINEERING CHEMISTRY
2313
ENGINEERING AND PROCESS DEVELOPMENT 0.04
>
c
0.00
t,
hechemical plant must be kept' low if this process is to succeed, The design of an efficient and economical chemical separation process is thus one of the most critically import,ant factors in making a success of nuclear breeding. Fast Breeder. We have seen that it is possible to use plutonium and uranium-238 in a breeding reactor, provided it is of the fast t,ype. We have also seen that. under these conditions it is not desirable to dilute the fuel very much with ot'her materials. For this reason fast breeders, such a s the experimental breeder react,or, usually consist of a reactor core in which concentrated fuel is burned, and a reactor blanket in which neutrons from the core are absorbed by fertile materials to produce fissionable mat.erial. Such a fast reactor flowsheet for the plutonium-uranium238 cycle is shown in Figure 5 , The consumption of uranium238 here is down to 445 grams per day, as in the thermal thorium hreeding cycle. The process is more complicated, however, because of the need for t M - 0 separat,ion plants. One of these works on blanket materials t,o separate plutonium bred there and send it to thecore, and to recover the unreacted uranium-238 and ret,urn this to the blanket. The second separation plant removes fission pr0duct.s from the core material and recycles it for another pass through the core. The recycle quantities in this flowsheet are strictly hypot,hetical and are given primarily to indicat,e that, high recovery efficiency will be required in both of these separat'ion plilnt,~. Now that the nuclear feasibility of breeding in a fast
INDUSTRIAL AND ENGINEERING CHEMISTRY
Vol. 45, No. 11
I&EC LECTURE SERIES-Nuclear
i
reactor has been demonstrated by the experimental breeder reactor, the economic feasibility of this process will depend to a very large extent on the development of sufficiently cheap and efficient chemical processes for recovering material from the blanket and core. I n concluding the discussion of these breeding processes, it may be noted that the raw fuel cost is a completely negligible component in the cost of power, even assuming that uranium-238 and thorium cost $35 a pound, which is most unlikely, as they are less reactive than normal uranium. The contribution of raw fuel to power costs in these flow sheets is only 0.01 mil per kilowatt hour. . Summary of Discussion of Fuel Cycles. Comparison of these fuel cycles shows that, roughly speaking, the more one spends for chemical processing of spent fuel, the lower will be the raw fuel component in the cost of nuclear power. Much more development work will be required before it is possible to say where the economic balance lies. From the discussion of fuel cycles, it can be seen that the main chemical engineering processes of importance in nuclear power plants are (1) separation of isotopes, notably uranium-235 and deuterium, (2) the concentration of uranium and thorium from their ores, (3) the purification of uranium and thorium for rehctoi use, and (4) separation of reactor products.
Power
trated uranium-235 for weapons can be easily modified for production of the slightly enriched uranium-235 required in the thermal reactor fuel cycle previously described. Consequently, it seems probable that this phase of the nuclear power industry will remain for some time in government hands and that the price for slightly enriched uranium will be set by the Government.
Isotope Separation
We have seen that the two types of thermal reactor of greatest present interest are the reactor using normal uranium moderated with heavy water, or a reactor using slightly enriched uranium and not requiring deuterium. Thus, for thermal, power-producing reactors some form of isotope separation process is now required. I n foreign countries where plants for the production of uranium-235 have not been constructed, the heavy water route is a popular one. I n this country, where a large uranium-235 isotope separation capacity has been installed by the Atomic Energy Commission, the use of slightly enriched uranium-235 now looks more attractive. I n the end the choice between these two isotope separation alternatives will depend on the relative costs of the two processes. Uranium-235 Separation. The process which has been adopted in this country for concentration of uranium-235.k the gaseous diffusion process. This is a highly classified topic about which little can be said today, particularly with reference to costs. It is obvious, however, that a plant designed to produce concen-
Plutonium-Producing Reactors a t Hanfor d Works, Atomic Energy Commission
Deuterium. Technology of deuterium separation is much less hemmed in by security restrictions than the separation of uranium-235. The recent report (3) of the Commonwealth Edison Co. and Public Service Co. of Northern Illinois on nuclear power reactor technology has given the unit price of heavy water as $82 per pound, and tlie inventory requirements of a 211.5-Mw. power plant as 250 tons, worth $41,000,000. This amounts to $200 per kilowatt generating capacity, which is more than the total cost of a conventional SPENT FUEL power plant. It is apparent that at this unit cost, KG, / D. *lo heavy water is a very expensive reactor material. REACTOR FUEL U-235 0.148 0.32 There is therefore a strong incentive t o develop -KG. /Q. V o U - 2 3 8 45.163 more economical processes than those now in use. U-235 0 . 4 6 1.0 PU 0 . 2 4 4 0.53 Table I1 lists five processes which might be used U-238 4 5 . 5 4 , , F. P. 0.445 0.97 for the separation of deuterium. The three distilNATURAL U lation processes, being the most familiar, will be REACTOR , taken up first. The distillation of water was one KG, /D. O/o of the two methods of preparing heavy water used U-235 0.563 0.72 by the Manhattan District. Because of the low U-238 77.637 I relative volatility (1.017) for this separation. the p.L ISOTOPE HEAT 1 4 0 0 M W process has the disadvantage of requiring a high reflux ratio and a large number of plates. The rePLANT sult is that the heat consumption per unit of outPLANT put is very large and a large number of towers must 25% EFF'Y be used for even a moderate production rate. To DEPLETED U be economic, this process needs a very cheap KC, /Q, 3 source of heat and must use column intervals which U - 2 3 5 0.103 0.32! combine high capacity with low holdup and low U - 2 3 8 32.097 pressure drop. This process should benefit mate32.2 rially from further chemical engineering developFigure 2. Once-through Flow of Slightly Enriched Uranium Fuel ments.
--
--
November 1953
' ; 1
INDUSTRIAL AND ENGINEERING CHEMISTRY
2375
ENGINEERING AND PROCESS DEVELOPMENT Distillation of ammonia has the advantage of a higher relative volatility than water but the disadvantages of requiring operation at reduced temperatures and of using a substance that is not available in unlimited quantities as water is. The result is that the maximum production rate of heavy water which could be obtained from the distillation of ammonia depends on the amount of ammonia produced commercially. As an example, a plant making 600 tons of ammonia per day could produce a maximum of 58 tons of heavy water per year. A single power-producing reactor may require as much as 250 tons of heavy water.
Table 11.
Deuterium Separation Processes
Exchange Reactions
HzO H2O
+ HD + HDO + H P + DCI $ HDO + HC1
Distillation HoO - HDO NHs - NHzD H2 - HD
Equilibrium Constant a t 25' C. 3 70 4 69
S . B. Pt.,
O
100 - 33.4 -252.7
C.
Vapor Pressure Ratio 1.017 1.037 1.73
1% deuterium from natural steam in three exhange Mactors and one electrolytic cell. The process as here represented is economic only in case theie is a profitable use for the hydrogen exhausted from the plant. This is the case of Trail and in Norway, where the hydrogen is used for ammonia synthesis. Even so, this process is worth considering only where electric power is very cheap. Developments which would make this process more economical would be the discovery of some way of carrying out this exchange reaction in the liquid phase, so that it could be carried out in a single multistage contactor, and development of means for reversibly combining hydrogen and oxygen to recover the electric current required to run the cells. The exchange reaction between steam and hydrogen chloride has two advantages over steam and hydrogen: (I) the more faorable equilibrium constant and (2) the fact that the reaction vtakes place in liquid water without requiring catalysis. Consequently, this process could be run as a countercurrent operation in a multistage gas-liquid contactor. It has the serious disadvantage of using an extremely corrosive mixture. As far as known, it has not been employed commercially. Production of Uranium and Thorium
Sources of Uranium. The principal sourcee of uranium are The distillation of hydrogen has a much more favorable relative listed in Table I11 in order of decreasing uranium content. The volatility than either water or ammonia, but requires operation at approximate tonnages of uranium reserves available in each a temperature within a few degrees of the absolute zero. Still, material as estimated by Bain ( 1 ) are also given. Reserves in the use of liquid hydrogen as a rocket propellant and for other purthe uranium ores, pitchblende and carnotite. are small compared poses has provided industrial experience in operating at these with those potentially available in South African gold ore and low temperatures, so that distillation of hydrogen is worth serious certain bituminous shales. The uranium content of the latter consideration as a source of deuterium. It is subject to the same materials is very low, of the order of a few ounces per ton, and it sort of limitation as distillation of ammonia, in that the amount of is here that the principal chemical engineering problems assocideuterium which can be produced this way is limited to the ated with the extraction of uranium will be found. In Bain's amount of hydrogen which has other industrial uses. It would not opinion, the largest ready source of industrial uranium is the gold pay to produce hydrogen solely for the extraction of deuterium. deposits of South Africa. This is particularly true of ore which My own opinion is that of all the processes listed in this table, the can be worked profitably for its gold content alone, so that the distillation of hydrogen will make heavy water a t the lowest cost, feed to the uranium extraction operation will already have been and could be used advantageously to the extent that hydrogen mined and crushed, with the principal material handling costs made for other purposes is available as feed for such a plant. paid for by the gold. Plans have already been made to recover Each of the exchange reactions makes use of an equilibrium uranium from the tailings of 13 gold mines. between water and a hydrogen-containing gas in which deuteThe extraction of uranium from Florida phosphate rock is anrium concentrates selectively in the water component of the mixother example of a profitable by-product operation. Processes ture. The steam-hydrogen reaction was used at Trail, British have been developed for recovering uranium from fert(i1iaerplants; Columbia, for the production of some of the Manhattan Project's heavy water, and is being used by the NorskHydro Co. in Norway for the production of the heavy water going into European reactors. The steam-hydrogen reaction requires catalysis SPENT FUEL and, so far, this has been effected only in the gas KG. /D O/o phase. A schematic flowsheet for this process is PLUTONIUM U - 2 3 5 0.014 0 . 0 6 7 shown in Figure 6. The equipment consists of a 136G, / D U-23820.755 series of catalytic exchange reactors connected in PU 0 . 1 3 6 0.635 Each stage is fed with countercurrent cascade. F.P. 0 . 4 4 5 2.08 ! steam enriched in deuterium from the next higher REACTOR FUEL 21.35 CHEMICAL stage and hydrogen depleted in deuterium from the - REACTOR next lower stage, The mixture is brought into exchange equilibrium and cooled to condense the water. The water is passed on t o the next lower REACTOR HEAT I 4 0 0 M W stage as reflux, and the uncondensed hydrogen is I WASTE passed on to the next higher stage as vapor. 1 1 NATURAL U Reboiled hydrogen at the bottom of the cascade is provided by electrolyzing all of the water leavU-235 0.154 0.72 2 5 a b EFF'Y ing the last stage except that taken off as product. U - 2 3 8 21.196 Since electrolysis is also a selective operation for 21.35 POWER 100 MW concentrating hydrogen in the vapor and deuterium in the liquid, this electrolytic cell provides additional fractionation. As this flowsheet shows, it Figure 3. Flowsheet for Plutonium Recycle with Natural Uranium Fuel should be possible to produce water containing
fI Ii
*
I
i
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2376
INDUSTRIAL A N D ENGINEERING CHEMISTRY
Vel. 45, No. 11
I&EC LECTURE SERIES-Nuclear
REACTOR
SPENT
KG. /D U-233 TH 1
0.7
( 1.52OIo)
45.3 46.0
I
THORIUM
-
R*EACTOR
the Blockaon Chemical Co. is producing u r a n i u m from this source, and other companies are planning production (9). T h e r e c o v e r y of uranium from the enormous reserves available in bituminous shales is an even more difficult engineering undertaking, because in most instances these shales would be mined primarily for their uranium values, with only by-product credit from the fuel and sulfur which might also be obtained from them. This is a problem to challenge the mining engineers and chemical engineers of the atomic age. It w&s estimated in 1947 that i t would be possible t o produce
November 1953
F. P.
0.445
0.700
44.855 48.0
CHEMICAL
SE PAR AT 10N ,
nEAT
4 4 5 Gh
c
FUEL
U-233 TH
t
9l
$27 per pound. If this low price can actually be obtained, we need have no fear of exhausting OUT supply of nuclear fuels available a t reasonable costs. A 100-Mw. nuclear power plant using a breeding fuel cycle will consume only 0.17 ton of uranium per year. Clearly, if the nuclear and chemical problems of the breeder can be solved economically and if uranium can be extracted from shale for $27 or even more per pound, there will be no lack of cheap fuel for nuclear power plants. Some selective extraction process as successful as cyanidization of gold ores is needed for recovery of uranium from these low grade materials. The reagent should combine only with the uranium and
KG. / D
FEED
I I
4 0 0 MW
PLANT
FISSION PRODUCTS
-
PLUTONIUM 0.447 KG,
U-238
U-238 O.445KGb
445 G.
I
POWER
-
REACTOR BLANKET
U-238
, NEUTRONS
I
/D.
44 253 KG, / D
44.698 KG, /D.
~
ID.
Power
I
I
0/0
-
pu
1.9 G I . / D .
BLANKET SEP. P L A N T FISSION PRODUCTS
0,445 FG, /D.
c
-
HEAT/ 400 MW
POWER
Figure 5.
\
104 M W
Flowsheet of Fast Plutonium-Uranium-238 Breeder
INDUSTRIAL AND ENGINEERING CHEMISTRY
2371
ENGINEERING AND PROCESS DEVELOPMENT
National Reactor Testing Station, Idaho
Thorium. Thorium has been less extensively prospected for than uranium, and less jnformation is available about its occurrence and extraction. The main source of thorium is thought to be monazite sand, which is a heavy mineral consisting of rare earth phosphates containing from 5 t,o 8% of thorium. Very large deposit's of this mineral are found in beach sands on t,he coasts of India and Brazil, but the total reserves do not appear to be as great, as those of uranium in shale. Perhaps u-hen lowgrade deposits of thorium are sought, the reserves of this element mill be found to be as great as those of uranium a t corresponding concentrat,ions. Processes generally similar to those used for purifying uranium may be used for thorium. Separation of Reactor Products
Isotopes Present. Spent fuel from a nuclear reactor contaiiis a large assortment of different chemical elements and isotopes
because of the variety of ways in which t,he fuel may undergo fission. Fission products contain all the elements in the periodic table between germanium and gadolinium. Some of these are short-lived and decay rapidly, but a baker's dozen or more need to be considered when designing processes for separation of reactor products. The most important neutron-absorbing and longlived radioactive isotopes in irradiated uranium are listed in Table V. Objectives of Separation. The most important objective in processing spent fuel from a nuclear reactor is recovery 01 the fissionable material, plutonium or uranium-233. A second objective in some instances is recovery of the fertile material, uranium or thorium, for recycle to the reactor core or blanket. Most of these materials must be treated for removal of neutronabsorbing fission products. In addition, it is usually necessary to remove radioactive fission products from materials to be recycled to the reactor, so that reactor fuel elements can subsequently he processed without hazard. Other objectives in the processing of spent react,or fuel may include the isolation of individual fission products for subsequent industrial use, such as cesium-137, a valuable source of gamma radiation, and finally, storage of waste radioactive materials under safe conditions. Difficulties. Processing of spent reactor fuels is made peculiarly difficult by their intense radioactivity. This requires t,hat t,he
2378
Table IV.
Production of Pure Uranium Metal from Pitchblende
Step
Form of Uranium
Waste
1. Digest with HNOI 2. Add H2SOd
U02(N01)2 in HzO UOz(N0s)z in HzO
Gangue PbSO4, RaROI, and other insol. sulfates Fe, &In, B, ctc., in HzO
U O ~ ( N O Iin) ~ether Extract with diethyl ether 4. Wash with N20 U O ~ ( N O I )In Z Hz0 5. Ppt. U with H102 C04.2HzO 6. Heat UOC UOa 7. Reduce with Hz UOZ 8. React with HF gas UF4 9. Reduce with Ca U metal ingot metal in bomb 3.
Steps 1 to 2 Steps 3 to 9
Table V.
€LO HLO CaFi
U. S. Patent 2,506,945 (7) Process of French AEC (6)
Important Isotopes in Irradiated Uranium
Heavy Elements Uranium 235, 236, 238 Plutonium 239 Seutron-Absorbing Fission Products Ruthenium 103 Xenon 131, 135 Neodymium 143 Samarium 149, 151 Europium 151, 152, 155 Gadolinium 155
INDUSTRIAL AND ENGINEERING CHEMISTRY
Long-Lived Radioactive Fission Products 85 Krypton 89, 90 Strontium 91 Yttrium 93,95 Zirconium 95 Niobium 99 Technetium 103, 106 Ruthenium 129, 131 Iodine 133 Xenon 137 Cesium Barium 140 141, 144 Cerium Praseodymium 143 147 Neodymium 147 Promethium
Vol. 45, No. 11
ILEC LECTURE SERIES-Nuclwr P o w u process equipment be surrounded by massive shielding, and that provision be made to remove the substantial amounts of heat which are asaociated with this radioactivity, and in some instances to cope with dewmpsition of solvents and construction materials from the radiations emitted by the materials being prooassed. Another difficulty is the crib i d mass haaard, which is pmnent whenever fiesionable material is handled at s u b s h t i d concentrations. This often requires a limitation in the sise of batches being p d , or in the dimensions of individual pieces of equipment. A third difficulty is the high degree of recovery which is usually required b* e w ~ eof the great value of the h i o n &le materialsb e i p d . A fourth d e p x of separation specified is the high .~ for the removal of radioactive fiasion products, whose wncentration must &xnetimes be reduced by a factor.of l O , ~ , o o O . Another source of difficulty which must already be obvious is the large number of components preaent with elements of such diverse properties ae the alkali cesium, and the man-made elements technetium (resembling mange nese) and Promethium (one Of the Inre earths).
WASTE HYDROGEN
1 9850 MOLES
I
0.005Vo D
I
FEED STEAM
, NO. 3
I
T-2
-_________________ OXYGEN 4925 MOLES
ELECT. CELL
PRODUCT WATER Figure 6.
10.122%
-
100 MOLES I -10 D
Flowsheet of Steam-Hydrogen Exchange Process for Ddutariurn Oxide
A h a 1 difficulty, and one which wae not originally anticipated, is the chemiesl similarity between urnnium and plutonium. These elementa are aa similar chemically ae two members of the rare earth series, F d novel and special methods had to be developed to eepafate them. Processes Used. Among the processes which have been proposed for separating the components of spent reactor fuels may be mentioned carrier precipitation (a p m e s s which waS used by the Manhattan Project), distillation of volatile halides, ion exchange, and solvent extraction. We will choose the last of these aa an example of a separation process which may be used for reactor products.
.
Solvent Extraction Procsas. The flow sheet of Figure 7 has been proposed by Bruce (e) for the separation of phtonium, uranium, and &ion producte from irradiated uranium metal. This process takes advantage of the extractability of bexavalent uranium and hexavalent plutonium by diethyl ether, and of the greater esse of reduction of plutonium from valence 6 to valence 4 compared with uranium. Irradiated metal is dissolved in nitric acid and the uranium and plutonium are odded to the hexavalent state. The principal fission pmducte to be separated from these elements are represented aa rare earth trinitrates. The aqueous solution is extraeted in a countercurrent column with a light phase wnsisting of diethyl ether. Fission products are washed out of the ether extract with 8 scrub solution consisting of e e . . dium nitrate in water. The extract from the &st column is paasedtoasewnd column where theplutonium is back-extraoted into the water phase by a m b solution containing a reducing agent capable of reducing the plutonium but not the uranium. The ether extract from the 8-nd column pseses to a third column, where the uranium is back-eatracted with water.
Integntion of Reactor with Fuel Processing plant The
Rgure 7. " M e e t af S d W M Fxfiis&w~ 'Proteft'fiir Irra&8%8'utbhRhh
November l&
$2)
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INDUSTRIAL A N D ENGINEERING CHEMISTRY
pmoess just d d h e d has the disadvantage that the metallic reactor fuel is dissolved in wid, and that before the uranium and plutonium can be re-uaed
23l9
ENGINEERING AND PROCESS DEVELOPMENT and a given separation can be cariied out in a much lower column height than is needed for more conventional extraction equipment. h second type of contactor is the mixer-settler, which consists of a series of solvent extraction cells connected horizontally in countercurrent cascade with mixing in each cell. Each cell contains a mixing compartment, a settling compmtment for disengaging the individual phases, and overflow weirs to direct the flow of the two phases countercurrently to adjacent cells. A recent device of this kind is the Pump-Mix mixer settler developed by Coplan, Davison, and Zebroski of the Knolls Atomic Powri Laboratory, General Electric Co. (4). ii third new type of solvent extractor is a mix-and-settle e\tractor developed by M. R. Fenske of Pennsylvania State College working under contract to the Standard Oil Development Co.; this extractor is to be manufactured and sold by Kational Rereaich C‘orp. It consists of a vertical stack of solvent extraction stages, each of which contains a mixing compartment and a settling compartment. A vertical. reciprocating rod passes through all of the mixing compartments and provides stirring in each stage. Overflow channels between adjacent stages provide counterflouof light and heavy phases and gaskets and seals prevent unwanted bypassing. The compactness of this extractor and the simple stirring mechanism are particular advantages. All of these extractors are valuable for radiochemical processes because they provide many stages of separation in a relatively compact space, and thus reduce the volume of shielding required, which is one of the big costs in a radiochemical separation plant. These compact, multistage extractors should also be useful in other inorganic separations and in processing pharmaceuticals and fine chemicals. Liquid Metal Heat-Transfer Equipment a t Knolls Atomic Power Laboratory
in a reactor, they must be reduced again t o metal m d fabricated into fuel elements. A notable simplification would result if the fuel burned in the reactor and processed in the separation plant were of the same chemical form, and even better, if the fuel could be pumped as liquid from one unit to the other. Development of such an integrated, fluid-fuel reactor and separation plant is being carried out by the Oak Ridge National Laboratory and the Dow-Detroit Edison group. It is an especially challenging chemical engineering development. Inorganic Solvent Extraction
Applications. The process just described for the separation of uranium and plutonium from spent reactor fuels is illustrative of the renaissance of solvent extraction as applied to inorganic chemical processes. Other applications which have been developed in the atomic energy program include the separation of the various rare earth elements and the separation of hafnium from zirccnium. These separations have been greatly aided by the intelligent use of complexing and salting agents in the aqueous phase and chelating agents in the organic phase. The versatility of these agents is so great that there are few elements other than the alkalies and alkaline earths which cannot be extracted in this way. Other possible fields of application include the separation of manganese from iron, nickel from cobalt, gold from silver, and the various platinum group elements. Types of Equipment. Along with the discovery of valuable reagents for use in solvent extraction has come the development of improved types of solvent extraction contacting equipment. A few of these improved types may be mentioned. The pulse column, originally proposed by van Dijck (8) hai proved to be well suited for inorganic solvent extraction (6). It consists of a vertical sieve-plate column in which a hydraulic pulse is superimposed upon the normal countercurrent flow of the two phases. Dispersion of one phase in the other is increased,
2380
Conclusion
Successful development of economic nuclear power may follow several different routes, and each of these has its own chemical engineering problems. The problems discussed include the production of heavy water; extraction of uranium from its ores, notably abundant, lo^ grade deposits such as uranium-bearing shales; production of pure uranium metal; and recovery of uranium, plutonium, and thorium from the complex mixture of unfamiliar and radioactive elements present in spent reactor fuel. Development of workable processes and reduction of cost to the extent required by the economics of nuclear power can be achieved only by large scale experimental development work and by actual production experience. Chemists and chemical engineers have much to contribute t o this program. This program will give rise to developments outside of the nuclear power field; improved solvent extraction equipment is one example. Literature Cited (1) Bain, G. W., Econ. Geol., 45, 274-321 (1950). (2) Bruce, F. R., “Chemical Development of Radiochemical Processes,” U. S. Atomic Energy Commission, AECD-3449 (1952). (3) Commonwealth Edison Co. and Public Service Co. of Northern
Illinois; Dow Chemical Co. and Detroit Edison Co.; Monsanto Chemical Co., and Union Electric Co.; and Pacific Gas and Electric Co. and Bechtel Corp., “Nuclear Power Reactor Technology,” U. S. Atomic Energy Commission, May 1953. (4) Coplan, B. V., Davison, J. K., and Zebroski, E. L., U. S. Patent 2,646,346 (1953). ( 5 ) Eichner, C., Goldschmidt, B., and Vertes, P., Bull. SOC. chint.
(France), (5) 18, 140 (1951). (6) Griffith, Jasney, and Tupper, S. M., thesis in chemical engineering, Massachusetts Institute of Technology, 1952. (7) Thomas, H. C., and Tomcufcik, A. S., U. S. Patent 2,506,945 (May 1950). (8) Van Dijck, W. J. D., U. S. Patent 2,011,186 (1935) (9) U. S. Atomic Energy Commission, Semiannual Report, JanuaryJune 1953. (10) Zinn, W. H., A’ucZeo?aics, 10, 8 (September 1952). RECEITED for reiriew September 15, 1953.
INDUSTRIAL AND ENGINEERING CHEMISTRY
ACCEPTED October 15. 1953.
Vol. 45, No. 11
