2.1 235 U

Table 12.2 lists methods that have been used on an industrial or large pilot-plant scale to enrich uranium in 235 U.

Method	Status
Gaseous diffusion of UF6	Three large plants operating in United States; large plants operating in the Soviet Union and China; smaller plants operating in England and France; large plant being constructed in France
Centrifugation of UF6	Large pilot plants operating and commercial plants under construction in England and Holland; large plant to be built in United States
Thermal diffusion of UF6	Small amount of slightly enriched UF6 produced in United States in 1945; process abandoned
Electromagnetic separation of UCI4	Used in United States in 1945 for first large-scale production of highly enriched 235 U; process abandoned in 1946
Separation nozzle process	Process demonstrated on large pilot-plant scale at Karlsruhe, Germany; semicommercial plant being built in Brazil
UCOR process	Process demonstrated in pilot plant at Valindaba, Union of South Africa; commercial plant under consideration

Table 12.2 Methods for enriching335 U

_УІ( 1 — y) “ */(!-*)

W*»UFt mjssUF6

Gaseous diffusion process. Figure 12.1 illustrates the principle of one stage of the gaseous diffusion process. Stage feed gas, UF6, flows past a diffusion barrier made of porous material with very fine holes, smaller than the mean free path of the UF6 molecules. About half of the feed gas flows through the barrier to a lower-pressure region. The gas passing through the barrier is slightly richer in 235 U than the gas remaining on the high-pressure side, because the mean speed of 235 UF6 molecules is slightly higher than that of 238 UF6 molecules. These mean speeds are in the inverse ratio of the square roots of the molecular weights of the two molecules. Under practical operating conditions the ratio of 235 UF6 atoms to 238 UF6 atoms in the enriched UF6 fraction passing through the barrier, yj(l —у), to the corresponding ratio in the depleted UF6 fraction remaining behind, x/(l — x), is in the ratio of their mean speeds:

The ratio [v/( 1 —y)]/[x/( 1 —x)] is called the stage separation factor and is denoted by a. Analogous separation factors are used to characterize all separation processes. A value of a close to unity indicates that the separation is difficult; a value far from unity, easier. For gaseous diffusion of UF6, a is so close to unity that the process must be repeated many times for a useful degree of separation. To do this, the low-pressure enriched UF6 must be recompressed to the feed pressure and cooled. The depleted UF6, which experiences some pressure loss, must also be recompressed (not shown).

Because of the small change in enrichment from a single stage, for a useful degree of enrichment, it is necessary to use many stages in series in countercurrent cascade. Figure 12.2 shows how stages are connected together in such a cascade. On each stage a motor-driven compressor takes partially depleted gas from the next higher stage and partially enriched gas from the next lower stage and recompresses them before passage through a cooler and the diffusion barrier. To separate natural uranium feed containing 0.00711 fraction 235 U into product containing 0.03 and tails 0.002 fraction 23SU requires 1272 stages. The minimum total

interstage flow in such a cascade is obtained when the compositions of the streams mixed at each point A are equal. Such a cascade is called an ideal cascade. In such a cascade, the interstage flow rate M from a stage where the 235 U fraction is у is

2РІУР-У)

(а — 1>(1 — У)

where P is the flow rate of product containing yp fraction 235 U. The theory of such an ideal cascade is developed later in this chapter, and details of the gaseous diffusion process are given in Chap. 14.

Figure 123 is a photograph of the large gaseous diffusion plant of the U. S. Department of Energy at Portsmouth, Ohio, which use 4080 stages to enrich 233U to 97 percent.

The gas centrifuge. Figure. 12.4 shows the principle of the type of countercurrent gas centrifuge proposed 20 years ago by the German engineer, Gernot Zippe [Zl], and now generally adopted by groups continuing development of this promising method of isotope separation. Such a centrifuge consists of a rapidly rotating cylindrical bowl made of a material with high strength-to-density ratio. The UF6 gas rotating inside in this bowl is subjected to centrifugal accelerations thousands of times greater than gravity. This makes the pressure at the outer radius of the bowl millions of times greater than at the axis and causes the concentration of 238 UF6 relative to 235 UF6 to be appreciably higher at the outer radius than at the axis. In a machine made of fiberglass running at the highest speed possible without mechanical failure, the 233 U content at the center of the bowl can be as much as 18 percent higher than at the

Figure 12.4 Zippe gas centrifuge sche­matic.

outside. In addition, longitudinal countercurrent flow of UF6 is induced by a system of rotating baffles and stationary scoops. In Fig. 12.4, gas enriched in 235UF6 at the center flows downward and gas enriched in 238 UF6 at the outside flows upward. Under these conditions the gas toward the bottom of the bowl becomes progressively richer in 235 UF6 and the gas at the top richer in 238 UF6. By making the bowl sufficiently long, the concentration difference between top and bottom can be made many times greater than between center and outside.

Gas centrifuges of greater capacity than described by Zippe have been developed in the United States, England, Germany, and Holland. Commercial centrifuge plants are operating in England and Holland and are planned in the United States. The power consumption of the centrifuge process is much lower than gaseous diffusion, and it is expected that separation costs will become lower. The process is described in more detail in Chap. 14.

Thermal diffusion of UF6. The thermal diffusion process makes use of the small difference in 23SU/238U ratio that is established when heat flows through a mixture of 23SUF6 and 238 UF6. The principle of the process is described in Chap. 14. The process was used [Al] in 1945 in the United States by the Manhattan Project to enrich uranium to 0.86 percent 235U. This slightly enriched material was used as feed for an electromagnetic separation plant. Although the process could be put into production quickly because of the simplicity of the equipment, it
was very inefficient, with very high heat consumption per unit of output. Consequently, when the more efficient gaseous diffusion plant came into operation at Oak Ridge, the thermal diffusion plant was dismantled. Thermal diffusion is a useful method, however, for separating small amounts of isotopes for research purposes. It is used, for example, at the Mound Laboratory to enrich 13C from 90 to 99 percent.

Electromagnetic processes. The possibility of using electromagnetic means for separating isotopes was established by Thomson [T5] in 1911. When Thomson passed a beam of positive neon ions through electric and magnetic fields, two traces were produced on a photographic plate, one for 20 Ne and the other for 22 Ne. The modem mass spectrometer works on the same general principle. With it, the existence of naturally occurring isotopes of 61 elements has been established, and isotopic abundances and masses have been determined (App. C).

In 1940, Nier and co-workers [N2] used a mass spectrometer to separate around 0.01 /Jg of

235 U from 238 U, to show that 235 U was the fissionable isotope of uranium. Because of its demonstrated ability to separate 235 U, the electromagnetic method was the first one selected by the Manhattan District for large-scale production of this isotope [S5]. Under the direction of Lawrence [LI ] at the University of California, mass spectrometers of greatly increased capacity were developed. The end result was the calutron[43] electromagnetic isotope separator used in the Y-12 plant at Oak Ridge, in which, in 1944, the first kilograms of 235U were produced.

When the gaseous diffusion plant came into operation, the cost of separating MSU electromagnetically was found to be higher, and in 1946, the Y-12 plant was taken off uranium-isotope separation. Some of this equipment is now being used to produce gram quantities of partially separated isotopes of most of the other polyisotopic elements, for research uses. These units have also been used to separate artificially produced isotopes, such as

236 U from irradiated uranium, and the various plutonium isotopes.

Large-capacity electromagnetic isotope separation equipment has also been developed in Russia [Z2], and at Harwell [S4], Amsterdam [K2], and other centers of nuclear research

[K5] •

Becker separation nozzle process. Recently there has been increased interest in aerodynamic processes in which partial separation of isotopes is obtained in flowing gas streams subjected to high linear or centrifugal acceleration. The aerodynamic process about which most information is available is the Becker separation nozzle process.* This originally employed linear acceleration of UF« through a divergent nozzle, but now uses a combination of linear and centrifugal acceleration through a curved slit.

Figure 12.5 is a cross section of the slit-shaped separation element used in the most fully tested form of the Becker nozzle process. Feed gas consists of a mixture of about 5 m/o (mole percent) UF6 and 95 m/o hydrogen at a pressure of around 1 atm. This flows into a low-pressure region through a long curved slit, or “nozzle” (perpendicular to the plane of the figure), with first a convergent, then a divergent cross section. The change in cross section accelerates the gas mixture to supersonic speed, and the curved groove downstream of the slit produces a centrifugal field. This sets up a concentration gradient in the mixture, with the gas adjacent to the curved wall enriched in 238 U relative to 235 U. A knife-edge downstream from the slit divides the stream into a more-deflected light fraction and a less-deflected heavy fraction.

Dilution of UF6 with hydrogen has two beneficial effects. The mixture has a much higher sonic velocity than pure UF6, so that much higher flow velocities are attainable, and inert gas makes the isotope separation factor greater than would be predicted for the prevailing centrifugal field. A separation factor of 1.015 can be obtained with a mixture of 5 percent UF6-95 percent H2 flowing through a pressure ratio of 3.5.

A more complete description of the process is given in Chap. 14. A semicommercial plant using this process is being built in Brazil.

UCOR process. The UCOR process, developed by the Uranium Enrichment Corporation of South Africa, Ltd., also makes use of high-speed flow of UF6-hydrogen mixtures through sharply curved ducts. By using a new cascade technique, called the Helikon, in which an axial-flow compressor handles several streams simultaneously without mixing, it is expected that natural uranium can be enriched to 3 percent 235 U with from 90 to 115 multistage compressor modules. A partial description of a South African pilot plant using this process was given by Roux and Grant [R2].

Laser-based processes. In addition to the processes listed in Table 12.2, intensive research is being conducted on using high-intensity, tunable lasers to separate uranium isotopes by making use of the small differences in absorption spectra of 235 U and 238 U or one of their compounds. Laser-based processes have demonstrated capability for selective separation of isotopes of many elements on a small scale and are considered promising candidates for eventual large-scale economic production of enriched uranium. Letokhov and Moore [L3] provide a good review of laser isotope separation work through 1976.

2.2 Deuterium

Commercial production of deuterium has been almost universally in the form of heavy water, DjO. Table 12.3 lists processes that have been used for production of heavy water at rates above a ton per year. These processes are divided into two classes. Parasitic processes take feed

from a primary plant producing hydrogen or ammonia synthesis gas (75 percent H2, 25 percent N2), extract deuterium from it, and return the depleted hydrogen for commercial use, usually ammonia synthesis. Self-contained processes have heavy water as their sole product and use natural water as feed. Generally speaking, the parasitic processes produce heavy water at lower cost, but their output is limited to the deuterium contained in the feed gas, which seldom contains more than 0.013 a/о (atom percent) deuterium. Even with complete deuterium extraction, a large plant producing 1000 short tons (t) of ammonia synthesis gas per day and operating 330 days/year could yield only

/10001 NH3 /330 days / 3 atoms H /0,00013 atoms D /20tD2Q / day / yr / molecule NH3/ atom H /t-molD20//

2 atoms 1

Concentration of deuterium by the electrolysis of water was proposed by Washburn and Urey [Wl], used by Lewis [L4] to make the first samples of pure D20, and employed for the first production of heavy water on a large industrial scale by the Norsk Hydro Company, at Rjukan, Norway. The Rjukan plant makes use of cheap hydroelectric power to produce electrolytic hydrogen for ammonia synthesis and by-product heavy water.

When Germany occupied Norway in World War II, this plant was producing 1.5 MT/year of heavy water, and around 90,000 MT/year of ammonia. The water being electrolyzed contained 21 MT/year of heavy water, of which 10 could have been recovered by burning hydrogen enriched in deuterium from the higher stages of the plant and recycling the deuterium-rich water. This, however, would have reduced the ammonia output by 23/300 MT/year. The German scientists Harteck, Hoyer, and Suess [C2] conceived the ingenious idea of recovering deuterium from the hydrogen gas by absorption in water, by making use of the exchange reaction

HD + H2O^HDO + H2 A= 3.0

in which deuterium concentrates in the water. A nickel catalyst for carrying out this reaction in the gas phase was developed. One catalytic reactor was installed at Rjukan, and others to bring the heavy-water production up to 5 MT/year were planned, but the plant was destroyed in 1943 in a series of daring commando raids. It was rebuilt after the war and has been in operation since then.

At about the same time, a similar exchange process was developed by Urey and Taylor [M5, S2, Tl], working under the Manhattan Project in the United States. The Standard Oil Development Company designed the exchange equipment [Bl] and installed it in the electrolytic hydrogen plant of the Consolidated Mining and Smelting Company, at Trail, British Columbia, where it was operated until 1955. This plant produced 6 MT D20/year at a concen­tration of 237 w/o (weight percent) D20. Final concentration to 99.7 w/o D20 was by elec­trolysis. The cost was $ 130/kg D20.

A second method for the industrial production of heavy water, used by the Manhattan Project in the United States [M5], was the distillation of water. Three plants having a total capacity of 13 MT D20/year were built at Army Ordinance plants. Because the relative volatility for separating H2 О from НЕЮ is only 1.03 at atmospheric pressure, the size of equipment and the heat consumption of these plants per unit of D2 О produced was very high, and the cost of heavy water was greater than in other processes. Nevertheless, the distillation of water was attractive as a wartime production method because the process needed little development work and used standard equipment. These plants were shut down after the war. More recently, distillation of water has come to be one of the most satisfactory methods for final concentration of heavy water.

Because the relative volatility for separation of deuterium by the distillation of liquid hydrogen is around 1.5 at atmospheric pressure, the size and heat consumption of a hydrogen distillation plant would be much smaller than that of a water distillation plant producing the same amount of deuterium. Plants to concentrate deuterium by the distillation of liquid hydrogen were designed by German engineers [C2] and by the Manhattan Project [M5] during World War II, and by Hydrocarbon Research, Inc. [H6], in the United States, but none of these plants was built because of uncertainty about the performance of industrial equipment operating at the very low temperatures needed to liquefy hydrogen. In 1949 a group of Soviet engineers undertook the development work necessary to ensure success of this type of plant, and in 1958 announced [Ml] that a plant producing deuterium by distillation of electrolytic hydrogen had been in operation in the Soviet Union for some years. The plant consists of multiple units, each with a capacity of around 4 MT D2 O/year.

In 1958, two companies specializing in cryogenic engineering put into operation experi­mental plants for concentrating deuterium by distillation of ammonia synthesis gas (75 percent H2, 25 percent N2). Socifte de Г Air Liquide designed and built one at the ammonia plant of Office National Industriel de l’Azote (ONIA), at Toulouse, France, which is operated by Compagnie Fran$aise de l’Eau Lourde, jointly owned by Air Liquide and ONIA. Gesellschaft fur Iinde’s Eismaschinen designed and built a second deuterium plant at the ammonia plant of Farbwerke Hoechst, at Hoechst, Germany. The production rates of the plants were roughly 2 and 6 MT D2 О/year, respectively. Because of the small size of these plants, the high local cost of electric power, and the less-than-natural deuterium content of the available synthesis gas, the cost of heavy water produced in these plants was high. After sufficient information had been obtained to permit design of larger plants at other locations where local conditions were more favorable, both plants were shut down in 1960.

In 1959, Sulzer Brothers designed and built a plant to distill electrolytic hydrogen enriched to six times the natural abundance of deuterium, which was available at the ammonia plant of Emswerke AG, at Ems, Switzerland [HI]. At this plant, about 2 MT/year of heavy water were

produced at a cost near $62/kg. The cost at Ems was lower than at Toulouse or Hoechst because of the higher deuterium content of feed and the low content of nitrogen and other condensable impurities in electrolytic hydrogen. This plant has been shut down because production of the electrolytic hydrogen that fed the heavy water plant has become too costly. In 1961, a 14 MT/year plant of this type was built by Linde to distill electrolytic hydrogen enriched to three times the natural abundance of deuterium, which was available at the Indian government’s ammonia plant at Nangal, India.

Another process that has been used to extract deuterium from ammonia synthesis gas is the deuterium-exchange reaction between liquid ammonia and gaseous hydrogen:

NH3(I) + HDfc) — NH2D(0 + HjGO

In the presence of potassium amide, KNH2, as catalyst dissolved in liquid ammonia, equilibrium favors concentration of deuterium in the liquid phase. A 26 MT/year plant using this process was operated at Mazingarbe, France, in the late 1960s, and three larger plants with a combined capacity over 200 MT/year are being built in India.

All of the previously mentioned plants except those employing distillation of water were parasitic to a synthetic ammonia plant. Their deuterium-production rate is limited by the amount of deuterium in ammonia synthesis gas. To produce heavy water at a sufficient rate, a growing industry of heavy-water reactors requires a deuterium-containing feed available in even greater quantity than ammonia synthesis gas. Of the possible candidates, water, natural gas, and petroleum hydrocarbons, water is the only one for which an economic process has been devised, and the dual-temperature hydrogen sulfide-water exchange process is the most economic of the processes that have been developed.

This process, invented by Spevack [S7] and developed independently by Geib [C2] in Germany, makes use of the fact that the separation factor a for exchange of deuterium between liquid water and gaseous hydrogen sulfide,

H2 0(/) + HDSfe) — HDO(0 + H2 ЗД is etc = 2.32 at 32°C and ah = 1.80 at 138°C

By running liquid water countercurrent to recycled gaseous hydrogen sulfide through first a cold tower and then a hot tower, as shown schematically in Fig. 12.6, water enriched in deuterium may be withdrawn from the water leaving the cold tower. The principle of the process and process flow sheets are described in detail in Chap. 13.

The first plant of this type, designed by the Girdler Corporation and operated by E. I. du Pont de Nemours and Company, built at the Wabash Ordnance Plant at Dana, Indiana, in 1952 but later shut down, gave this process the name the G-S process, for Girdler-Sulfide. Three improved units, each with a capacity of 160 MT/year, were designed, built, and operated by du Pont at Aiken, South Carolina [B2]; one is still in operation at a reduced capacity of 69 MT/year.

Figure 12.7 is a photograph of this plant. The world’s principal heavy-water production capacity is found in Canada, where G-S plants with a total capacity of 4000 MT/year are in operation or under construction.

2.3 Lithium Isotopes

Many methods have been used to achieve partial separation of lithium isotopes on a small scale. Examples of processes and reported separations are listed in Table 12.4. A process somewhat similar to the last one listed in this table, involving countercurrent exchange of lithium isotopes between aqueous lithium hydroxide and lithium amalgam, is to be used in a plant being built

Figure 12.6 Dual-temperature water-hydrogen sulfide process.

by Eagle Kcher Industries, Inc., at Quapaw, Oklahoma, to produce 1000 kg 99.99 percent 7 Li per year at an approximate price of $3/g.

2.4 10 В

Table 12.5 compares four processes that have been used for concentrating 10B. The research that led to the first commercial production of 10 В was carried out by Crist and Kirshenbaum [C5] in the laboratory of H. C. Urey at Columbia University in 1943. As reported-by Kilpatrick and co-workers [Kl], it was concluded that the most satisfactory process consisted in the equilibrium distillation of the complex of boron trifluoride and dimethyl ether, BF3-(СНз^О. When this substance vaporizes, it dissociates partially according to the reaction

BF3 -(СН3)а О — BF3 + (СН3)г О

The isotopic exchange equilibrium

10BF3C?) + UBF3 -(CH3)20(1) * nBF3(g) + I0BF3 ’(СН3)20(Г)

is then established, with an equilibrium constant of 1.027 at 100°C [К2]. When the liquid is distilled at 100°C, the vapor phase is 60 percent dissociated.

(1.027X0.6) + (1.000X0.4) = 1.016 (12.4)

This value has been confirmed experimentally [Kl].

A semicommercial plant based on this process was built and operated for the Manhattan Project by the Standard Oil Company of Indiana [C4]. In 1953, the U. S. Atomic Energy Commission authorized construction of a larger plant at Niagara Falls, New York, with the Hooker Electrochemical Company as operating contractor [Ш]. This plant produced 460 kg/year of 10B at an enrichment of 92 a/о 10B. The plant was shut down in January 1958. Eagle Picher Industries, Inc., has been producing 10В at Quapaw, Oklahoma, by this process since 1973 and is expanding capacity to 1000 kg/year. The cost is from $5 to $15/g.

A plant producing 2 kg of 10 В per year by equilibrium distillation of the complex of BF3 and diethyl ether, BF3 *(Сг H5 X O, was operated by 20th Century Electronics, Ltd., in New Addington, England [Е1]. The process, developed by the U. K. Atomic Energy Authority (UKAEA), is generally similar to the U. S. process using the dimethyl ether complex. Both plants are operated at subatmospheric pressure, to minimize irreversible decomposition of the complex.

Distillation of BF3 is another process that has been used to concentrate I0B. This has the advantage over the processes using ether complexes of BF3 that decomposition is not a problem, so that the plant can be operated at atmospheric pressure and can be scaled up without special concern about increased column pressure drop. Disadvantages of BF3, however, are that the separation factor is only 1.0075 [Nl], and the reflux condenser must be operated

Figure 12.7 Heavy-water plant at Aiken, South Carolina. (Courtesy of U. S. Energy Research and Development Administration.)

Method	Investigated by	Reference	Separation factor	Enrichment obtained
Differential ion migration
Fused LiCl	Klemm et al.	[K4]		7 Li to 97%; 6 Li to 16%
	Klemm	[КЗ]		7 Li to 99.974%
Fused LiBr	Lunden	[L7]
Fused LiNC>3	Vallet et al.	[VI]
Electrolysis of LiCl in H2 О	Johnston and Hutchison	[J2]	1.055
	Perret et al.	[PI]	1.05-1.07
Molecular distillation of Li	Trauger et aL	[T6]	1.06	6 Li to 9% in 8 stages
Equilibrium distillation of Li	Perret et al.	[PI]	1.03
Chemical exchange
Li amalgam vs. LiCl in alcohol	Lewis and MacDonald	[L5]		‘Lito 14%
Li amalgam vs. LiBr in DMFt	Perret et al.	[PI]	~1.05
Ion exchange
Aqueous LiCl vs. zeolite	Taylor and Urey	[T3]	1.022
Aqueous LiCl vs. zeolite	Sessions et al.	[S3]	1.004-1.006
Aqueous LiCl vs. Dowex 50 X 12	Perret et al.	[PI]	1.002	6 Li to 10.2%
Aqueous LiCl vs. Dowex 50	Lee and Begun	[L2]	1.0038
Chemical exchange between lithium
amalgam and aqueous solution
of lithium compound	Saito and Dirian	[SI]

+ DMF, dimethyl formamide.

Reference	Method of separation	Working substance	Operating conditions	Separation factor	10 В production rate, kg/year	Percent10 В
Pressure, Torr	Temperature, °С
[М3]	Distillation + exchange	bf3-(ch3)2o	150-275	91-104	1.016	460	92
[El]	Distillation + exchange	bf3-(c2h5)2o	20-53	10-75	1.016	2	95
[Nl]	Distillation	BF3	760	-101	1.0075	26.5	95
[H2]	Distillation + exchange	BF3*anisole		25	1.032	—	—

at temperatures in the inconvenient range between the melting point of BF3 (—127°C) and its normal boiling point (—101°C). Despite these difficulties, the process was used successfully in the Soviet Union [M4] to produce O. S kg/year of 10 В enriched to 83 percent, and in England by the UKAEA [Nl] to produce 26.5 kg/year enriched to 95 percent. I0B concentrates in the liquid phase, as in the exchange equilibrium.

2.5 13C

Natural carbon contains 1.11 percent 13 C. This isotope was first produced commercially at a rate of around 1 g/day by the Eastman Kodak Company [S8], using the exchange reaction between HCN gas and NaCN solution developed in 1940 by Urey and co-workers [H5]. The separation factor is 1.013.

13 C has also been produced by the low-temperature distillation of carbon monoxide, in a process developed by London and co-workers [Jl, L6]. A carbon monoxide distillation plant has been in operation at Harwell since 1949, producing 0.4 g/day of 13 C at 60 to 70 percent enrichment. Simultaneously, the plant produces 0.045 g/day of 18 О at 5 to 6 percent enrichment. The separation factors for these two separations are

12C160/13C160: 1.011

12CI60/12C180: 1.008

A carbon monoxide distillation plant at Los Alamos Scientific Laboratory produces 4 kg 13C/year [A2] at 90 percent enrichment.

2.6 15 N

Natural nitrogen contains 0.365 percent 15 N. Methods that have been used for separating 15 N on a small scale are listed in Table 12.6.

The exchange reaction between NH3 gas and NH4NO3 in aqueous solution was used by Thode and Urey in 1939 to obtain the first samples of enriched 1SN, and was employed by the Eastman Kodak Company to produce 15N at a rate of around 1 g/day. The only production of ls N in the United States at present is by distillation of NO at Los Alamos [М2].