NUCLEAR WEAPONS TECHNOLOGY 2

II-5-9
Figure 5.0-2. Nuclear Weapons Foreign Technology Assessment Summary
Legend: Sufficient Technologies Capabilities: ¨¨¨¨ exceeds sufficient level ¨¨¨ sufficient level ¨¨ some ¨ limited
Because two or more countries have the same number of diamonds does not mean that their capabilities are the same. An absence of diamonds in countries of concern may
indicate an absence of information, not of capability. The absence of a country from this list may indicate an absence of information, not capability.
Country Sec 5.1
Enrichment
Feedstocks
Production
Sec 5.2
Uranium
Enrichment
Processes
Sec 5.3
Nuclear
Fission
Reactors
Sec 5.4
Plutonium
Extraction
(Reprocessing)
Sec 5.5
Lithium
Production
Sec 5.6
Nuclear
Weapons
Design
and
Development
Sec 5.7
Safing,
Arming,
Fuzing,
and Firing
Sec 5.8
Radiological
Weapons
Sec 5.9
Manufacturing
of
Nuclear
Components
Sec 5.10
Nuclear
Weapons
Development
Testing
Sec 5.11
Nuclear
Weapons
Custody,
Transport,
and
Control
Sec 5.12
Heavy
Water
Production
Sec 5.13
Tritium
Production
Argentina ¨ ¨¨ ¨¨ ¨ ¨¨ ¨¨¨¨ ¨¨
Austria ¨ ¨ ¨ ¨ ¨¨
Belgium ¨¨¨¨ ¨ ¨ ¨¨
Brazil ¨¨ ¨¨ ¨¨ ¨ ¨¨ ¨ ¨¨
Canada ¨¨¨ ¨¨¨¨ ¨¨¨ ¨ ¨ ¨ ¨¨¨¨ ¨¨¨¨
China ¨¨¨ ¨¨¨ ¨¨¨ ¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨
Czech Republic ¨¨
France ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨
Germany ¨¨¨ ¨¨¨¨ ¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨ ¨¨¨ ¨¨¨
India ¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨ ¨¨ ¨¨ ¨¨¨ ¨¨¨¨ ¨¨¨¨
Iran ¨ ¨¨ ¨¨ ¨ ¨¨
Iraq ¨¨ ¨ ¨¨¨ ¨ ¨¨¨ ¨ ¨¨ ¨ ¨¨
Italy ¨¨¨ ¨¨ ¨¨ ¨¨ ¨¨
Japan ¨¨ ¨¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨ ¨¨ ¨¨
Netherlands ¨¨¨ ¨¨ ¨¨ ¨¨ ¨¨ ¨
North Korea ¨¨ ¨¨ ¨ ¨¨¨ ¨¨ ¨¨
Pakistan ¨¨ ¨¨¨ ¨¨ ¨¨¨¨ ¨ ¨¨ ¨¨ ¨¨ ¨
Russia ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨¨¨
South Africa ¨¨¨¨ ¨¨ ¨¨¨ ¨¨¨ ¨¨ ¨ ¨¨¨¨ ¨¨ ¨¨ ¨¨
South Korea ¨¨¨¨ ¨¨¨ ¨ ¨ ¨
Sweden ¨¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨¨ ¨¨ ¨¨¨ ¨¨
Switzerland ¨¨¨¨ ¨¨¨ ¨¨¨¨ ¨¨¨ ¨¨ ¨¨¨ ¨¨
Taiwan ¨¨¨ ¨¨¨ ¨¨¨ ¨ ¨
Ukraine ¨¨¨ ¨ ¨ ¨¨ ¨
United Kingdom ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨
United States ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨ ¨¨¨¨

 

 

II-5-10
OVERVIEW
This subsection covers technologies utilized in the conversion of uranium ore
concentrates to highly purified uranium hexafluoride (UF6) and uranium tetrachloride
(UCl4) for subsequent use as feedstock in a uranium-enrichment process. Gaseous
UF6 is used as the feed in the gas centrifuge and gaseous diffusion processes, and UCl4
is used as feed in the electromagnetic isotope separation (EMIS) process.
Uranium ore concentrates, also known as yellowcake, typically contain 60–
80 percent uranium and up to 20 percent extraneous impurities. There are two commercial
processes used to produce purified UF6 from yellowcake. The primary difference
between the two processes—solvent extraction/fluorination (“wet process”) and
fluorination/fractionation (“dry process”)—is whether the uranium is purified by solvent
extraction before conversion to UF6 or by fractional distillation of the UF6 after
conversion.
In the wet process, yellowcake is dissolved in nitric acid (HNO3), and the insoluble
residue is removed by filtration or centrifugation. Uranium is separated from
the acid solution with liquid-liquid extraction, the uranyl nitrate product is decomposed
to uranium trioxide (UO3) via thermal denitration, and the trioxide is reduced to
uranium dioxide (UO2) with hydrogen or cracked ammonia (NH3). In most cases, the
standard Purex process, using tri-n-butyl phosphate (TBP) in a hydrocarbon diluent,
separates uranium from its impurities in the extraction step.
In the dry process, the conversion and purification steps occur throughout the
process. If the yellowcake was produced by the alkali-leach process (yields Na2U2O7),
the sodium must be removed from the material by partial digestion in sulfuric acid
followed by ammonia precipitation of ammonium diuranate [(NH4)2U2O7]. The ammonium-
containing uranium salt is decomposed to UO3 by heating, and this oxide is
reduced to UO2 with hydrogen or cracked NH3.
The remaining steps used to produce UF6 for both processes are similar in that the
UO2 is converted to UF4 by hydrofluorination (using hydrogen fluoride gas—HF).
The UF4 (impure in the dry process) is converted to UF6 using electrolytically generated
fluorine gas (F2). In the dry process, the UF6 is purified in a two-stage distillation
step. Direct fluorination of UO3 to UF6 has been used, but this procedure is more
amenable to relatively small capacity plants.
The EMIS uranium-enrichment process uses UCl4 for its feed material. Uranium
tetrachloride is produced by the reaction of carbon tetrachloride (CCl4) with pure UO2
at 700 °F.
RATIONALE
A country choosing to join the nuclear weapons community must acquire the necessary
weapons (fissile) material (235U or 239Pu). A state selecting uranium for its weapons
must obtain a supply of uranium ore and construct an enrichment plant because the
235U content in natural uranium is over two orders of magnitude lower than that found
in weapons grade uranium (>90 percent 235U). Nearly all uranium enrichment plants
utilize UF6 as their feed. A country may select the EMIS process, which uses UCl4 as
its feed material, for enriching uranium.
FOREIGN TECHNOLOGY ASSESSMENT (See Figure 5.0-2)
The processes outlined above are unclassified and have been described extensively
in the literature on the nuclear fuel cycle. Many countries around the world
have extracted uranium from its ores or from yellowcake. The processes for preparing
the feedstocks are basic industrial chemistry.
The enabling technologies are those which use HF, NH3, F2, CCL4, and precursor
uranium compounds to prepare UF6 and UCL4.
SECTION 5.1—ENRICHMENT FEEDSTOCKS PRODUCTION
Highlights
• UF6
and UCl4 are the principal compounds used as inputs to
uranium enrichment processes.
• Manufacture of these feedstocks is straightforward industrial
chemistry.
• These processes are unclassified and widely known.

 

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Table 5.1-1. Enrichment Feedstocks Production Technology Parameters
Technology
Sufficient Technology
Level
Export Control
Reference
Critical
Materials
Unique Test, Production,
and Inspection Equipment
Unique Software
and Parameters
Purification of yellowcake
(wet process)
Knowledge of liquid-liquid
extraction systems
Experience in using HNO3
NTL 8F;
NRC J
Yellowcake
Nitric acid (HNO3)
tri-n-butyl phosphate
(TBP)
Refined kerosene
Filters; centrifuges;
pulse columns; concentration/
thermal denitration
systems; tanks
resistant to HNO3
Distribution coefficients
for many elements
Aqueous solubility for
many compounds
Purification of yellowcake
(dry process:
produces impure UO2)
Ability to handle H2 at
elevated temperature
NTL 8F;
NRC J
Yellowcake (should
not contain high
concentrations of
sodium or
magnesium)
H2SO4
See citations below
Furnace; air filtration
equipment; fluidized bed;
temperature control;
heat exchangers
None identified
UO2 preparation Ability to handle H2 at
elevated temperature
NTL 8F;
NRC J
H2
NH3
Moving bed reactor;
rotary kiln; air filtration
equipment; fluidized bed;
temperature control
system
None identified
UF4 preparation Ability to manage HF at
elevated temperature
Ability to provide a dry
environment
NTL 8F;
NRC J
HF Stirred fluidized bed
reactors; rotary kiln;
moving bed/screw
reactor; air cleaning
equipment (filters,
scrubbers); fluorideresistant
equipment
None identified
UF6 preparation (used in
gaseous diffusion and
gas centrifuge
enrichment processes)
Capability to control
quantities of fluorine gas.
Ability to operate a flame
tower with F2.
Experience in removing H2
from electrolytic cells (F2
production) .
Experience in operating in an
anhydrous environment
NTL 8F;
NRC J
F2
HF
KF • 2HF
Flame tower reactor;
fluidized bed reactor;
condensers (cold traps);
electrolytic cells (for F2
production); highamperage,
low-voltage
supply (for F2 production);
air-cleaning
equipment; F2-resistant
equipment (Monel);
fluoride-resistant
equipment; UF6 storage
Careful temperature
control is required for
fluorination
UCl4 preparation (used in
EMIS enrichment
process)
Water-free environment must
be provided
NTL 8F;
NRC H
CCl4 Stirred fluidized bed
reactors; rotary kiln;
moving bed/screw
reactor; air-cleaning
equipment (filters,
scrubbers)
Reasonable control of
temperature

II-5-12
Table 5.1-2. Enrichment Feedstocks Production Reference Data
Technology Technical Issues Military Applications Alternative Technologies
Purification of yellowcake (wet
process)
HNO3 solutions are relatively
hazardous and require moderate care
in handling
None identified Direct fluorination of UO3
Purification of yellowcake (dry
process produces impure UO2)
H2 presents an explosive hazard None identified Direct fluorination of UO3
UO2 preparation H2 presents an explosive hazard None identified Step may be bypassed using
direct fluorination
UF4 preparation Inappropriate use of HF can present
health problems.
Improper operation of tower reactors
may cause plugging (caking).
None identified Step may be bypassed using
direct fluorination
UF6 preparation (used in gaseous
diffusion and gas centrifuge
enrichment processes)
Producing F2 is not an easy task.
Flame towers can be difficult to
operate.
Moisture-sensitive material difficult to
handle.
UF6 product is feed to most U
enrichment processes
None identified
UCl4 preparation (used in EMIS
enrichment process)
Moisture-sensitive material difficult to
handle
UCl4 product is feed to the EMIS
enrichment process
None identified

 

II-5-13
SECTION 5.2—URANIUM ENRICHMENT PROCESSES
OVERVIEW
It is generally recognized that the acquisition of fissile material in sufficient quantity
is the most formidable obstacle to the production of nuclear weapons. Fissile
material production consumes the vast majority of the technical, industrial, and financial
resources required to produce nuclear weapons. For example, production of fissile
materials—highly enriched uranium (HEU) and plutonium—accounted for more
than 80 percent of the $1.9 billion (1945 dollars) spent on the Manhattan Project.7
Fissile materials can produce energy by nuclear fission, either in nuclear reactors
or in nuclear weapons. The principal fissile materials of interest are 235U, 233U, and
239Pu. Uranium-235 is of particular interest because it is the only fissile material that
occurs in nature in significant quantity, and it can be used to construct a nuclear explosive
device if a sufficient quantity can be acquired. In a typical sample of natural
uranium, only 0.72 percent of the atoms are 235U atoms, and it can be assumed that all
of the remaining atoms are 238U atoms.8 Higher concentrations of 235U are required for
many applications, and the use of uranium isotope separation processes to increase the
assay of 235U above its natural value of 0.72 percent is called uranium enrichment.
While low-enriched uranium (LEU) could technically mean uranium with an assay
anywhere between slightly greater than natural (0.72 percent) and 20 percent 235U,
it most commonly is used to denote uranium with an assay suitable for use in a lightwater
nuclear reactor (i.e., an assay of <5 percent). Similarly, the term “highly enriched”
uranium (HEU) could be used to describe uranium with an assay >20 percent,
but it is commonly used to refer to uranium enriched to 90 percent 235U or higher (i.e.,
weapons-grade uranium). The term “oralloy” was used during World War II as a contraction
of “Oak Ridge alloy,” and it denoted uranium enriched to 93.5 percent 235U.
When plutonium is produced in a nuclear reactor, inevitably some 240Pu (as well
as heavier plutonium isotopes, including 241Pu and 242Pu) is produced along with the
more desirable 239Pu. The heavier isotope is not as readily fissionable, and it also
decays by spontaneous fission, producing unwanted background neutrons. Thus, nuclear
weapon designers prefer to work with plutonium containing less than 7 percent 240Pu.
A method for separating plutonium isotopes could be used to remove the heavier isotopes
of plutonium (e.g., 240Pu) from reactor-grade plutonium, thus producing nearly
pure 239Pu. Uranium isotope separation techniques [e.g., atomic vapor laser isotope
separation (AVLIS)] might be applied to this task. However, this would require mastery
of production reactor and reprocessing technologies (to produce and extract the
plutonium) in addition to isotope enrichment technology (to remove the heavier plutonium
isotopes). In practice, it is simpler to alter the reactor refueling cycle to reduce
the fraction of plutonium which is 240Pu.
Manhattan Project scientists and engineers explored several uranium-enrichment
technologies, and production plants employing three uranium-enrichment processes—
electromagnetic isotope separation (EMIS), liquid thermal diffusion, and gaseous diffusion
Ñwere constructed at Oak Ridge, Tennessee, during the period from 1943 to
1945. Centrifugation was tried, but the technology needed to spin a rotor at an appropriate
speed was not then practical on an industrial scale. The aerodynamic separation
processes developed in Germany and South Africa did not exist during World War II;
7 Richard G. Hewlett and Oscar E. Anderson, The New World: A History of the United States
Atomic Energy Commission, Volume 1, 1939/1946, University of California Press, a 1990
edition of a book originally published by Pennsylvania State University Press in 1962.
8 Natural uranium typically has a composition of 0.0055 atom % 234U, 0.7205 atom % 235U,
and 99.274 atom % 238U. For most purposes, the tiny fraction of 234U can be neglected.
Highlights
• The acquisition of fissile material in sufficient quantity is the most
formidable obstacle to the production of nuclear weapons.
• Gas centrifuges are today the technology of first choice for
enriching uranium, based on process economics and minimum
consumption of electricity.
• Technologies considered obsolete for commercial uranium
enrichment, such as electromagnetic isotope separation (EMIS), can
be employed by a proliferant state at some added cost in electric
power and labor requirements.
• Aerodynamic separation processes developed in South Africa and
Germany have proven satisfactory for a limited number of nuclear
weapons, despite their high cost to operate.
• Laser isotope separation (LIS) techniques are based on advanced
technologies and represent potential uranium enrichment processes
of the future.

 

II-5-14
neither, of course did laser isotope separation or plasma separation. The World War II
Japanese nuclear program made some attempts to find a purely chemical process.
RATIONALE
Methods of Separation
Electromagnetic Isotope Separation
The EMIS process is based on the same physical principle as that of a simple mass
spectrometer—that a charged particle will follow a circular trajectory when passing
through a uniform magnetic field. Two ions with the same kinetic energy and electrical
charge, but different masses (i.e., 235U+ and 238U+), will have different trajectories,
with the heavier 238U+ ion having the larger diameter. The different diameters of the
trajectories of the two uranium ions allow for the separation and collection of the
material in receivers or “collector pockets.” EMIS is a batch process that can produce
weapons-grade material from natural uranium in only two stages. However, hundreds
to thousands of units would be required to produce large quantities of HEU because of
the process’s relatively low product collection rate and the long cycle time required to
recover material between runs.
In the uranium EMIS process, uranium ions are generated within an evacuated
enclosure (called a “tank”) that is located in a strong magnetic field. For the EMIS ion
source, solid uranium tetrachloride (UCl4) is electrically heated to produce UCl4 vapor.
The UCl4 molecules are bombarded with electrons, producing U+ ions. The ions are
accelerated by an electrical potential to high speed and follow a circular trajectory in
the plane perpendicular to the magnetic field. In the U.S. EMIS separators, the ion
beam traverses a 180-deg arc before the ions pass through slit apertures at the collector.
A major problem with the EMIS process is that less than half of the UCl4 feed is
typically converted to the desired U+ ions, and less than half of the desired U+ ions are
actually collected. Recovery of unused material deposited on the interior surfaces of
the tanks is a laborious, time-consuming process that reduces the effective output of an
EMIS facility and requires a large material recycle operation.
In the U.S. EMIS program, production of weapons-grade uranium took place in
two enrichment stages, referred to as the a and b stages. The first (a) stage used
natural or slightly enriched uranium as feed and enriched it to 12–20% 235U. The
second (b) stage used the product of the (a) stage as feed and further enriched it to
weapons-grade uranium. To allow more efficient use of magnets and floor space, the
individual stages were arranged in continuous oval or rectangular arrays (called “racetracks”
or, simply, “tracks”) with separator tanks alternated with electromagnetic units.
The U.S. EMIS separators are referred to as “calutrons” because the development work
was carried out at the University of California (Berkeley) during the early 1940’s
using cyclotrons.
Although most applications of the EMIS process have been applied to the
commercial production of both stable and radioactive isotopes, all five recognized
weapons states have tested or used the EMIS process for uranium enrichment. Even
with the problems associated with using the process, an EMIS facility could be attractive
for a country desiring a limited weapons-grade uranium enrichment program. The
process might be especially appealing as a method for further enriching partially enriched
material. It has been well documented that EMIS was the principal process
pursued by the Iraqi uranium enrichment program. This occurred at a time when EMIS
had been discarded and largely forgotten as a method for uranium enrichment because
it is both energy intensive and labor intensive, and it is not economically competitive
with other enrichment technologies.
Thermal Diffusion
Thermal diffusion utilizes the transfer of heat across a thin liquid or gas to accomplish
isotope separation. By cooling a vertical film on one side and heating it on the
other side, the resultant convection currents will produce an upward flow along the hot
surface and a downward flow along the cold surface. Under these conditions, the
lighter 235U gas molecules will diffuse toward the hot surface, and the heavier 238U
molecules will diffuse toward the cold surface. These two diffusive motions combined
with the convection currents will cause the lighter 235U molecules to concentrate
at the top of the film and the heavier 238U molecules to concentrate at the bottom of the
film.
The thermal-diffusion process is characterized by its simplicity, low capital cost,
and high heat consumption. Thermal diffusion in liquid UF6 was used during World
War II to prepare feed material for the EMIS process. A production plant containing
2,100 columns (each approximately 15 meters long) was operated in Oak Ridge for
less than 1 year and provided a product assay of less than 1% 235U. Each of these
columns consisted of three tubes. Cooling water was circulated between the outer and
middle tubes, and the inner tube carried steam. The annular space between the inner
and middle tubes was filled with liquid UF6.
The thermal-diffusion plant in Oak Ridge was dismantled when the much more
energy-efficient (by a factor of 140) gaseous-diffusion plant began operation in the
1940’s. Today, thermal diffusion remains a practical process to separate isotopes of
noble gases (e.g., xenon) and other light isotopes (e.g., carbon) for research purposes.
Gaseous Diffusion
The gaseous-diffusion process has been highly developed and employed to produce
both HEU and commercial reactor-grade LEU. The United States first employed
gaseous diffusion during WWII and expanded its capacity after the war to produce
HEU. Since the late 1960’s, the U.S. facilities have been used primarily to produce
commercial LEU, with the last remaining HEU capacity being shut down in 1992.
China and France currently have operating diffusion plants. Russia’s enrichment
facilities have been converted from diffusion to centrifuge technology. Britain’s diffusion
facility was shut down and dismantled.

 

II-5-15
The gaseous-diffusion process depends on the separation effect arising from molecular
effusion (i.e., the flow of gas through small holes). on average, lighter gas
molecules travel faster than heavier gas molecules and consequently tend to collide
more often with the porous barrier material. Thus, lighter molecules are more likely to
enter the barrier pores than are heavier molecules. For UF6, the difference in velocities
between molecules containing 235U and 238U is small (0.4 percent), and, consequently,
the amount of separation achieved by a single stage of gaseous diffusion is small.
Therefore, many cascade stages are required to achieve even LEU assays.
The production of a sustainable, efficient separating membrane (barrier) is the
key to the successful operation of a diffusion plant. To obtain an efficient porous
barrier, the holes must be very small (on the order of one-millionth of an inch in diameter)
and of uniform size. The porosity of the barrier must be high to obtain high flow
rates through the barrier. The barrier must also be able to withstand years of operation
while exposed to corrosive UF6 gas. Typical materials for the barrier are nickel and
aluminum oxide.
Diffusion equipment tends to be rather large and consumes significant amounts of
energy. The main components of a single gaseous-diffusion stage are (1) a large cylindrical
vessel, called a diffuser or converter, that contains the barrier; (2) a compressor
used to compress the gas to the pressures needed for flow through the barrier; (3) an
electric motor to drive the compressor; (4) a heat exchanger to remove the heat of
compression; and (5) piping and valves for stage and interstage connections and process
control. The entire system must be essentially leak free, and the compressors
require special seals to prevent both out-leakage of UF6 and in-leakage of air. The
chemical corrosiveness of UF6 requires use of metals such as nickel or aluminum for
surfaces exposed to the gas (e.g., piping and compressors). In addition to the stage
equipment, auxiliary facilities for a gaseous-diffusion plant could include a large electrical
power distribution system, cooling towers to dissipate the waste process heat, a
fluorination facility, a steam plant, a barrier production plant, and a plant to produce
dry air and nitrogen.
Gaseous diffusion is unlikely to be the preferred technology of a proliferator due
to difficulties associated with making and maintaining a suitable barrier, large energy
consumption, the requirement for procuring large quantities of specialized stage equipment,
large in-process inventory requirements, and long equilibrium times.
Gas Centrifuge
The use of centrifugal fields for isotope separation was first suggested in 1919;
but efforts in this direction were unsuccessful until 1934, when J.W. Beams and coworkers
at the University of Virginia applied a vacuum ultracentrifuge to the separation
of chlorine isotopes. Although abandoned midway through the Manhattan Project,
the gas centrifuge uranium-enrichment process has been highly developed and used to
produce both HEU and LEU. It is likely to be the preferred technology of the future
due to its relatively low-energy consumption, short equilibrium time, and modular
design features.
In the gas centrifuge uranium-enrichment process, gaseous UF6 is fed into a cylindrical
rotor that spins at high speed inside an evacuated casing. Because the rotor
spins so rapidly, centrifugal force results in the gas occupying only a thin layer next to
the rotor wall, with the gas moving at approximately the speed of the wall. Centrifugal
force also causes the heavier 238UF6 molecules to tend to move closer to the wall than
the lighter 235UF6 molecules, thus partially separating the uranium isotopes. This separation
is increased by a relatively slow axial countercurrent flow of gas within the
centrifuge that concentrates enriched gas at one end and depleted gas at the other. This
flow can be driven mechanically by scoops and baffles or thermally by heating one of
the end caps.
The main subsystems of the centrifuge are (1) rotor and end caps; (2) top and
bottom bearing/suspension system; (3) electric motor and power supply (frequency
changer); (4) center post, scoops and baffles; (5) vacuum system; and (6) casing. Because
of the corrosive nature of UF6, all components that come in direct contact with
UF6 must be must be fabricated from, or lined with, corrosion-resistant materials.
The separative capacity of a single centrifuge increases with the length of the
rotor and the rotor wall speed. Consequently, centrifuges containing long, high-speed
rotors are the goal of centrifuge development programs (subject to mechanical constraints).
The primary limitation on rotor wall speed is the strength-to-weight ratio of the
rotor material. Suitable rotor materials include alloys of aluminum or titanium,
maraging steel, or composites reinforced by certain glass, aramid, or carbon fibers. At
present, maraging steel is the most popular rotor material for proliferants. With
maraging steel, the maximum rotor wall speed is approximately 500 m/s. Fiber-reinforced
composite rotors may achieve even higher speeds; however, the needed composite
technology is not within the grasp of many potential proliferants. Another limitation
on rotor speed is the lifetime of the bearings at either end of the rotor.
Rotor length is limited by the vibrations a rotor experiences as it spins. The rotors
can undergo vibrations similar to those of a guitar string, with characteristic frequencies
of vibration. Balancing of rotors to minimize their vibrations is especially critical
to avoid early failure of the bearing and suspension systems. Because perfect balancing
is not possible, the suspension system must be capable of damping some amount of
vibration.
One of the key components of a gas centrifuge enrichment plant is the power
supply (frequency converter) for the gas centrifuge machines. The power supply must
accept alternating current (ac) input at the 50- or 60-Hz line frequency available from
the electric power grid and provide an ac output at a much higher frequency (typically
600 Hz or more). The high-frequency output from the frequency changer is fed to the

 

II-5-16
high-speed gas centrifuge drive motors (the speed of an ac motor is proportional to the
frequency of the supplied current). The centrifuge power supplies must operate at
high efficiency, provide low harmonic distortion, and provide precise control of the
output frequency.
The casing is needed both to maintain a vacuum and to contain the rapidly spinning
components in the event of a failure. If the shrapnel from a single centrifuge
failure is not contained, a “domino effect” may result and destroy adjacent centrifuges.
A single casing may enclose one or several rotors.
Although the separation factors obtainable from a centrifuge are large compared
to gaseous diffusion, several cascade stages are still required to produce even LEU
material. Furthermore, the throughput of a single centrifuge is usually small, which
leads to rather small separative capacities for typical proliferator centrifuges. To be
able to produce only one weapon per year, several thousand centrifuges would be
required.
The electrical consumption of a gas centrifuge facility is much less than that of a
gaseous diffusion plant. Consequently, a centrifuge plant will not have the easily identified
electrical and cooling systems typically required by a gaseous diffusion plant.
Aerodynamic Processes
Aerodynamic uranium enrichment processes include the separation nozzle process
and the vortex tube separation process. These aerodynamic separation processes
depend upon diffusion driven by pressure gradients, as does the gas centrifuge. In
effect, aerodynamic processes can be considered as nonrotating centrifuges. Enhancement
of the centrifugal forces is achieved by dilution of UF6 with a carrier gas (i.e.,
hydrogen or helium). This achieves a much higher flow velocity for the gas than could
be obtained using pure UF6.
The separation nozzle process was developed by E.W. Becker and associates at
the Karlsruhe Nuclear Research Center in Germany. In this process, a mixture of
gaseous UF6 and H2 (or helium) is compressed and then directed along a curved wall at
high velocity. The heavier 238U-bearing molecules move preferentially out to the wall
relative to those containing 235U. At the end of the deflection, the gas jet is split by a
knife edge into a light fraction and a heavy fraction, which are withdrawn separately.
Economic considerations drive process designers to select separation nozzles with
physical dimensions as small as manufacturing technology will allow. The curved
wall of the nozzle may have a radius of curvature as small as 10 mm (0.0004 in.).
Production of these tiny nozzles by such processes as stacking photo-etched metal
foils is technically demanding.
A typical stage consists of a vertical cylindrical vessel containing the separation
elements, a cross piece for gas distribution, a gas cooler to remove the heat of compression,
and a centrifugal compressor driven by a electric motor.
The Uranium Enrichment Corporation of South Africa, Ltd. (UCOR) developed
and deployed its own aerodynamic process characterized as an “advanced vortex tube”
or “stationary-walled centrifuge” at the so called “Y” plant at Valindaba to produce
hundreds of kilograms of HEU. In this process, a mixture of UF6 and H2 is compressed
and enters a vortex tube tangentially at one end through nozzles or holes at velocities
close to the speed of sound. This tangential injection of gas results in a spiral or vortex
motion within the tube, and two gas streams are withdrawn at opposite ends of the
vortex tube. The spiral swirling flow decays downstream of the feed inlet due to
friction at the tube wall. Consequently, the inside diameter of the tube is typically
tapered to reduce the decay in the swirling flow velocity. This process is characterized
by a separating element with very small stage cut (ratio of product flow to feed flow)
of about 1/20 and high process-operating pressures.
Due to the very small cut of the vortex tube stages and the extremely difficult
piping requirements that would be necessary based on traditional methods of piping
stages together, the South Africans developed a cascade design technique, called
Helikon. In essence, the Helikon technique permits 20 separation stages to be combined
into one large module, and all 20 stages share a common pair of axial-flow
compressors. A basic requirement for the success of this method is that the axial-flow
compressors successfully transmit parallel streams of different isotopic compositions
without significant mixing. A typical Helikon module consists of a large cylindrical
steel vessel that houses a separating element assembly, two axial-flow compressors
(one mounted on each end), and two water-cooled heat exchangers.
For both of these aerodynamic processes, the high proportion of carrier gas required
in relation to UF6 process gas results in high specific-energy consumption and
substantial requirements for removal of waste heat.
Laser Isotope Separation
In the early 1970’s, significant work began on the development of laser isotope
separation technologies for uranium enrichment. Present systems for enrichment processes
using lasers fall into two categories: those in which the process medium is
atomic uranium vapor and those in which the process medium is the vapor of a uranium
compound. Common nomenclature for such processes include “first category—
atomic vapor laser isotope separation (AVLIS or SILVA)” and “second category—
molecular laser isotope separation (MLIS or MOLIS).”
The systems, equipment, and components for laser-enrichment plants embrace
(a) devices to feed uranium-metal vapor (for selective photoionization) or devices to
feed the vapor of a uranium compound (for photo-dissociation or chemical activation);
(b) devices to collect enriched and depleted uranium metal as product and tails in the
first category and devices to collect dissociated or reacted compounds as product and
unaffected material as tails in the second category; (c) process laser systems to
selectively excite the 235U species; and (d) feed preparation and product conversion

 

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equipment. The complexity of the spectroscopy of uranium atoms and compounds
may require incorporation of any number of available laser technologies.
AVLIS
The atomic vapor laser isotope separation (AVLIS) process is based on the fact
that 235U atoms and 238U atoms absorb light of different frequencies (or colors). Although
the absorption frequencies of these two isotopes differ only by a very small
amount (about one part in a million), the dye lasers used in AVLIS can be tuned so that
only the 235U atoms absorb the laser light. As the 235U atom absorbs the laser light, its
electrons are excited to a higher energy state. With the absorption of sufficient energy,
a 235U atom will eject an electron and become a positively charged ion. The 235U ions
may then be deflected by an electrostatic field to a product collector. The 238U atoms
remain neutral and pass through the product collector section and are deposited on a
tails collector.
The AVLIS process consists of a laser system and a separation system. The separator
system contains a vaporizer and a collector. In the vaporizer, metallic uranium is
melted and vaporized to form an atomic vapor stream. The vapor stream flows through
the collector, where it is illuminated by the precisely tuned laser light. The AVLIS
laser system is a pumped laser system comprised of one laser used to optically pump a
separate dye laser, which produces the light used in the separation process. Dye master
oscillator lasers provide precise laser beam frequency, timing, and quality control.
The laser light emerging from the dye master oscillator laser is increased in power by
passage through a dye laser amplifier. A total of three colors are used to ionize the 235U
atoms.
Many countries are pursuing some level of AVLIS research and/or development,
and major programs exist in the United States, France, Japan, and probably Russia.
Principal advantages of the AVLIS process include a high separation factor, low energy
consumption (approximately the same as the centrifuge process), and a small
volume of generated waste. However, no country has yet deployed an AVLIS process,
although several have demonstrated the capability to enrich uranium with the process.
While conceptually simple, the actual implementation of the process is likely to be
difficult and expensive, especially for countries with limited technical resources. The
AVLIS process requires much sophisticated hardware constructed of specialized materials
that must be capable of reliable operation for extended periods of time in a harsh
environment.
MLIS
The idea for the molecular laser isotope separation (MLIS) process was conceived
by a group of scientists at the Los Alamos National Laboratory in 1971. There are two
basic steps involved in the MLIS process. In the first step, UF6 is irradiated by an
infrared laser system operating near the 16 mm wavelength, which selectively excites
the 235UF6, leaving the 238UF6 relatively unexcited. In the second step, photons from a
second laser system (infrared or ultraviolet) preferentially dissociate the excited 235UF6
to form 235UF5 and free fluorine atoms. The 235UF5 formed from the dissociation precipitates
from the gas as a powder that can be filtered from the gas stream.
MLIS is a stagewise process, and each stage requires conversion of the enriched
UF5 product back to UF6 for further enrichment. CO2 lasers are suitable for exciting
the 235UF6 during the first step. A XeCl excimer laser producing ultraviolet light may
be suitable for the dissociation of 235UF6 during the second step. However, there is
currently no known MLIS optical system which has been successfully designed to
handle both infrared and ultraviolet. Consequently, most MLIS concepts use an all
infrared optical system.
In terms of the gas flow for the MLIS process, gaseous UF6 mixed with a carrier
gas and a scavenger gas is expanded through a supersonic nozzle that cools the gas to
low temperatures. Hydrogen or a noble gas are suitable as carriers. A scavenger gas
(such as methane) is used to capture the fluorine atoms that are released as a result of
the dissociation of 235UF6 molecules.
There are many complexities associated with the process, and the United States,
UK, France, and Germany have stated that their MLIS programs have been terminated.
Japan also has had a small MLIS program. South Africa has recently stated that
their MLIS program is ready to be deployed for low-enriched uranium (LEU) production.
Principal advantages of the MLIS process are its low power consumption and its
use of UF6 as its process gas.
Chemical and Ion Exchange
Chemical-exchange isotope separation requires segregation of two forms of an
element into separate but contacting streams. Since many contacts are required to
achieve the desired separation, the contacting process must be fast and achieve as
much separation as possible. For heavy elements such as uranium, achieving a suitable
separation factor involves contact between two valence (oxidation state) forms
such as hexavalent [U6+ as in uranyl chloride (UO2Cl2)] and the quadrivalent [U4+ as in
uranium tetrachloride (UCl4)]. The 235U isotope exhibits a slight preference for the
higher valence, for example, the hexavalent over the quadrivalent in the Asahi process
or the quadrivalent over the trivalent (U3+) in the French solvent-extraction process.
The chemical-exchange process, developed by the French, is commonly referred
to as CHEMEX. It uses the exchange reaction that takes place between two valence
states (U3+ and U4+) of uranium ions in aqueous solution. Isotopic enrichment results
from the tendency of 238U to concentrate in the U3+ compound while 235U concentrates
in the U4+ compound. It is therefore possible to obtain enriched uranium by removing
the U4+ ions with an organic solvent that is immiscible with the aqueous phase (concentrated
hydrochloric acid). Several possible extractants are available; however,
tributyl phosphate (TBP), the choice of the French, is typically used. TBP is diluted
with an aromatic solvent, and this organic phase moves countercurrent to the aqueous
phase through a series of pulsed columns.

 

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In the pulse column, the heavier aqueous phase is fed into the top of the column,
and the lighter organic phase is fed into the bottom of the column. A rapid reciprocating
motion is applied to the contents of the column, providing efficient and intimate
contact of the two phases. In an HEU plant, centrifugal contactors might be employed
particularly for the higher assay sections, since the stage times and corresponding specific
uranium inventory could be reduced significantly.
After passing through the column, the enriched and depleted uranium streams
must be chemically treated so that they can be recirculated through the column again
(refluxed) or sent to another column for additional enrichment. This requires complicated
refluxing equipment at both ends of the column.
The ion-exchange process was developed by the Asahi Chemical Company in
Japan and uses the chemical isotope effect between two valences (U4+ and U6+) of
uranium. In this process, the organic phase is replaced by a proprietary ion-exchange
resin. The aqueous phase flows through the stationary resin held in a column, and the
net effect of all the chemical reactions is a “band” of uranium that moves through the
ion-exchange column. The exchange between the unadsorbed uranium flowing through
the band and that adsorbed on the resin enhances the isotopic separation. In this continuous
separation system, 235U and 238U tend to accumulate respectively at the entrance
and exit ends of the adsorption band. In this process, it is economical to regenerate
many of the chemicals by reaction with oxygen and hydrogen in separate equipment.
The development and manufacture of the appropriate adsorbent beads are based
on technology and know-how gained by Asahi in over 25 years of ion-exchange membrane
development and manufacture. The adsorbent is a spherical bead of porous
anion-exchange resin with a very high separation efficiency and an exchange rate over
1,000 times faster than the rates obtained in most commercially available resins.
The two exchange processes discussed here are representative of exchange processes
now under study in several countries. At present, no country has built or operated
a full-scale uranium enrichment plant based on an exchange process. The primary
proliferation concern is that they are based on standard chemical engineering
technology (except for the proprietary ion-exchange resins).
Plasma Separation
The plasma separation process (PSP) has been studied as a potentially more efficient
uranium-enrichment technique that makes use of the advancing technologies in
superconducting magnets and plasma physics. In this process, the principle of ion
cyclotron resonance is used to selectively energize the 235U isotope in a plasma containing
235U and 238U ions. A feed plate of solid uranium serves as the source of neutral
uranium atoms. These atoms are vaporized by bombarding the plate with energetic
ions in a process called sputtering. A microwave antenna located in front of the plate
energizes free electrons which collide with neutral uranium atoms in the vapor
sputtering off the plate. This in turn displaces electrons from the uranium atoms and
produces a plasma of 235U and 238U ions.
The plasma is subjected to a uniform magnetic field along the axis of a cylindrical
vacuum chamber as the plasma flows from source to collector. The magnetic field is
produced by a superconducting magnet located around the outside of the chamber.
The high-strength magnetic field produces helical motions of the ions, with the lighter
235U ions spiraling faster and having a higher ion cyclotron frequency than the heavier
238U ions. As the ions move toward the collector, they pass through an electric field
produced by an excitation coil oscillating at the same frequency as the ion cyclotron
frequency of the 235U ions. This causes the helical orbit of the 235U ions to increase in
radius while having minimal effect on the orbit of the heavier 238U ions. The plasma
flows through a collector of closely spaced, parallel slats, the physical appearance of
which roughly resembles a venetian blind. The large-orbit 235U ions are more likely to
deposit on the slats, while the remaining plasma, depleted in 235U, accumulates on an
end plate of the collector. PSP is a batch process that would require several stages to
produce HEU from natural feed.
The only countries known to have had serious PSP experimental programs are the
United States and France. PSP became a part of DOE’s Advanced Isotope Separation
research and development program in 1976, but development was dropped in 1982
when AVLIS was chosen as the advanced technology of choice. The French developed
their own version of PSP, which they called RCI. Funding for RCI was drastically
reduced in 1986, and the program was suspended around 1990, although RCI is
still used for stable isotope separation.
Proliferation Implication Assessment
Uranium gun-assembled weapons are the easiest of all nuclear devices to design
and build. It is generally conceded to be impossible to prevent any nation having the
requisite amount of HEU from building one or more gun-assembled weapons. Therefore,
the acquisition of significant quantities of 235U or a facility in which to separate
the fissile material is an indicator that the acquiring state could be in the process of
gaining a rudimentary nuclear capability. Because HEU is used in certain research
reactors, another interpretation is possible. Because of the weapons potential, the United
States and France have sought to replace HEU-fueled reactors with ones using a lower
grade (<20% 235U, for example) of uranium which cannot be so readily converted to
weapons use. The uranium gun-bomb route was successfully taken by South Africa.
Any nation having uranium ore in sufficient quantity, a sufficiently well-developed
technological and industrial infrastructure, sufficient electric power, and the desire to
acquire nuclear weapons might well choose the uranium gun technology.
FOREIGN TECHNOLOGY ASSESSMENT (See Figure 5.0-2)
All five nuclear weapon states have demonstrated the ability to enrich uranium to
weapons grade. In addition, enrichment is a commercial process in The Netherlands

 

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and Japan. Germany has also demonstrated the ability to enrich uranium; the South
African nuclear weapons were made from 80–90% 235U produced indigenously. Brazil
and Argentina sought to build enrichment plants but have abandoned the effort.
Iraq used EMIS to enrich uranium prior to the Gulf War and was in the process of
building a centrifuge enrichment cascade. Iraq produced some enriched uranium (not
weapons grade) before the Gulf War terminated its program. Iran has invested large
sums in various enrichment schemes, some of which appear to have been clever scams
by outsiders, without achieving any significant enrichment capability. Pakistan has
built a gas centrifuge enrichment facility, believed to produce material for nuclear
weapons.
The nozzle enrichment process was to be used in Germany and in a plant to be
built in Brazil by NUCLEBRAS (a Brazilian firm) in cooperation with a German company,
Interatom. Neither plant appears to have been completed and placed in commercial
service.
Germany operates a commercial centrifuge enrichment plant for its nuclear power
industry. The Becker nozzle process is not believed to be in use anywhere in the world
today.