The document discusses the nuclear fuel cycle, including mining and processing of uranium ore, enrichment, fuel fabrication, use in reactors, and interim spent fuel storage. It provides details on various types of reactors, including boiling water reactors, pressurized water reactors, and fast breeder reactors. High temperature gas cooled reactors are also summarized. The full fuel cycle process and workings of different reactor designs are described in extensive technical detail over multiple paragraphs.

1. • Fuel Cycle
• How a Reactor Works
• High Temperature Gas Cooled Reactors
• Boiling Water Reactors
• Pressurized Water Reactors
• Fast Breeder Reactor
Fuel Cycle
The steps required to transform raw uranium ore into nuclear fuel are known collectively as the
fuel cycle. Uranium is a relatively abundant element in the earths crust. There are several
types of uranium ore, uranite and pitchblend are two forms of the same mineral, both are
uranium dioxide. Gummite is another common ore containing uranium oxide and water.
Orbernite is a rarer ore of uranium that usually does not occur in massive enough quantities to
make mining it economical but in Zaire it forms large masses. It is a hydrous copper uranium
phosphate. Aytunite is hydrous calcium uranium phosphate and is very abundant worldwide.
Carnotite is hydrous potassium uranium vanadate found in sandstone deposits in the
Southwestern United States and elsewhere. Uranophane is a somewhat less common mineral
found in several areas of the US, Saxony, Czechoslovakia and Zaire. It is hydrated calcium
uranium silicate.
According to Washington Nuclear Corporation “The percentage of U in uranium ores varies
widely: Namibian ore is about 0.035 percent U3O8, US ore ranges from 0.05 to 0.3 percent,
and in Northern Saskatchewan the U3O8 reaches 12.0 percent. Major deposits are also mined
in Australia and the former Soviet Union. Uranium is an abundant but not concentrated
material. Material mined in the various producing areas in 1997 amounted to 35,692 metric
tons, of that amount Canada produced 12,029 metric tons and the United States produced
2,170 metric tons. Canada produced 35% of the world’s production. Uranium may be mined in
several ways depending upon where the ore is located and how concentrated it is. Open pit
mining is used for deposits near the surface and underground mining is used for deeper high-
grade ore. In locations where the uranium ore is in sandstone below the water table of a
confined aquifer it may be mined by the in situ leach method by injecting a mildly alkaline
solution in to the rock to dissolve the uranium and then pumping it out.
The next step is milling, except for uranium mined by the insitu leach method which does not
require milling. The ore is brought out of the mine and ground finely, then a dilute solution of
acid or alkali is added to the powder to leach out the uranium. The resulting fluid is treated to
extract and concentrate the uranium. The liquor, as it’s called, is dried to produce a product
known as yellowcake. The yellowcake is 70 - 90 % Uranium oxide, U3O8. During conversion
the yellowcake is chemically converted to Uranium hexafluoride or Hex, UF6. This product is a
white solid which transforms into a gas at 56 C. There is only one Uranium Hexafluoride
Production Facility in the United States, Allied - Signal Inc. at Metropolis, Illinois. Hex is highly
toxic and corrosive so it must be handled carefully. It is approximately 99.28% Uranium238 and
0.71% Uranium235 this is the same as natural ore. U238 does not fission, therefore, the
percentage of U235 must be increased in order to sustain a chain reaction. In the United

2. States enrichment is done at two Gaseous Diffusion Plants, Paducah, Kentucky and
Portsmouth, Ohio. Both plants are operated by United States Enrichment Corporation, a
formerly government owned company which is now operated by a contractor. This process is
begun by heating the UF6 slightly to produce a gas which is them passed through a series of
diffusion barriers to increase the percentage of U235 to 3-5%. The gas must pass through
hundreds of semi permeable membranes with holes less than one millionth of a centimeter in
diameter. The membranes are in a series of pipes. The U235 is lighter and passes through the
barriers first. The U238 is heavier and passes more slowly through the barriers so it can be
captured and drawn off. Many sets of pipes are arranged in sets. Each set is known as a
cascade. The process uses a large amount of electricity and is not terribly efficient. The
enriched UF6 gas is cooled after the final cascade. The final product is solid UF6 with is then
converted into UO2 or Uranium Oxide. This power is compacted into pellets which are sintered
(heated) to produce ceramic UO2 pellets. These are loaded into tubes made from a Zirconium
and Tin alloy called Zircaloy or occasionally Stainless Steel. The air is removed and replaced
with Helium. The tubes are then sealed at each end with weld plugs. Sealing is intended to
contain fission products produced while in operation within the fuel rods. The tubes are loaded
into racks called assemblies and the finished assemblies are loaded into the reactor. The
individual tubes that make up the assembly are called rods. The number of fuel rods, their size
and the arrangement of the fuel assemblies varies according to the type of reactor. There are
seven commercial fuel fabrication plants in the United States, they are: ABB Combustion
Engineering in Hematite, Missouri; General Electric in Wilmington, North Carolina;
Westinghouse Electric in Columbia, South Carolina; Nuclear Fuel Services in Erwin,
Tennessee; Babcock and Wilcox Fuel Company in Lynchburg, Virginia; and Siemens Nuclear
Power Corporation in Richland, Washington.
The average Boiling Water Reactor assembly weighs 319.9kg, of which 183.3kg is Uranium,
measures 4.470m in length, and contains 63 fuel rods. The average Pressurized Water
Reactor assembly weighs 657.9kg, of which 461.4kg is Uranium, measures 4.059m in length,
and contains 264 fuel rods. The fuel assemblies are loaded into the reactor using a gantry
crane while the reactor is shut down. There may be 750 assemblies in a BWR, 150 in a PWR.
Of the 110 reactors operating in the United States, 73 are PWRs and 35 are BWRs. While the
reactor is operating the fuel reaches quite high temperatures. It tends to heat the Uranium
pellets unevenly. On the surface the pellets reach approximately 1200o F, while the core
reaches 2300o F. Both figures are substantially lower than the melting point of UO2 which is
5072o F but the differential heating tends to crack the pellets. The fuel assemblies are rotated
through the reactor, starting when first loaded on the outside and being rotated toward the
center later. The average lifespan of a fuel assembly is 4 to 6 years. Since the reactor must be
shutdown to load and unload fuel, utilities are changing from an annual cycle which requires 3
refuelings in a 3 year period, to an 18 month cycle which requires only 2 refuelings in the same
period. The fuel removed from the reactor is placed in a storage pool next to the reactor.
Currently there is no reprocessing of used commercial fuel rods in the United States, although
a number of European countries reprocess fuel from commercial plants.
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How a Reactor Works
A nuclear reactor consists of a containment vessel which surrounds the reactor vessel, a
number of Uranium fuel assemblies inside the reactor vessel, a loop of pipe that carries water

3. from the reactor to a steam generator and back to the reactor by means of a pump, another
loop of pipe to take steam from the steam generator to the turbine generator and then take
water back to the steam generator to be made into steam again by means of a pump. The key
to the process is the heat generated in the reactor by the fissioning of Uranium235. The reactor
is started by slowly withdrawing the control rods from the core to get the nuclear chain reaction
started. The fuel begins to fission, each atom of U235 that is struck by a free neutron in turn
produces free neutrons, which strike an atom of U235 to continue the chain reaction. The water
in the reactor acts as a moderator to slow the neutrons and make it more likely that they will
cause fissioning. The control rods can be moved in or out of the reactor to slow down or speed
up the fission reaction. The control rods contain material that absorbs neutrons, such as
cadmium or boron. When enough neutrons are absorbed, the reaction stops. In addition to
moderation the reaction the water acts as coolant to control the temperature of the core and
prevent the fuel from melting. The system operates under pressure, something like a kitchen
pressure cooker. This allows the water to reach much higher temperatures, nearly 300°F, than
it otherwise could without boiling. When this superheated water reaches the steam generator
the cool water in the secondary loop is immediately brought to a boil and converted into steam
to turn the blades of the turbine and generate electricity. This is the same principle that is used
in plants that burn oil, coal or gas, the heat is used to boil water and turn a turbine, the only
difference is the source of the heat. There are a number of different reactor designs in use, but
most of the reactors in the United States are either Boiling Water (BWR) or Pressurized Water
(PWR) reactors. Worldwide there are several other types including Gas Cooled, Pressurized
Heavy Water, Graphite and Water, Fast Breeders, Light Water Breeders, Magnox Gas Cooled.
Boiling Heavy Water, Light Water Cooled Heavy Water, and others. The various configurations
and moderator/coolant combinations all operate on the basic principal of heat produced by a
nuclear chain reaction being used to turn a turbine and generate power. There have been
individual examples of several of these other types of reactors built in the US. Some of these
have had operating problems or accidents and most are now closed.
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High Temperature Gas Cooled Reactors
Fort St. Vrain and Peach Bottom were high temperature gas cooled reactors, both are now
decommissioned . Peach Bottom was an experimental 40 megawatt reactor which went on line
in January of 1966 in Lancaster County Pennsylvania and closed in October of 1974. It was
operated by Peco Energy Company of Philadelphia. The Fort St. Vrain reactor was a 330
megawatt reactor ordered in 1965 from General Atomic Co., which went on line in December of
1973 in Platteville Colorado and closed in August of 1989. It was operated by Public Service
Company of Colorado. High temperature gas cooled reactors such as the Fort St. Vrain plant
are graphite moderated and cooled with helium. The helium coolant permits higher operating
temperatures of the fluid used, 760°–1430°F The steam generators of the plant produce steam
under pressure, 2400 psia and 1000°F Fort St. Vrain used a different type of fuel,
"microspheres of thorium and fully enriched uranium embedded in a carbonaceous binder" The
Thorium232 is transformed into Uranium233 in the core by neutron bombardment and the
U233 is fissionable, therefore the reactor could run longer before refueling. The plant shut
down prematurely due to "failed control rod drives and degradation of [the] steam generator
ring header".
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4. Boiling Water Reactors
Boiling Water Reactors are the second most common type in the United States, there are 35
operational BWRs at this time. The design of the BWR has two variants, Dual Cycle and Single
Cycle. In a dual cycle BWR there are two loops for the coolant, in a single cycle there is only
one. The vast majority of operating reactors are single cycle. The primary difference between
the two is that with a single cycle, radiation precautions must be taken around the turbines
since the water used to generate the steam used to turn the turbines has passed through the
core of the reactor. In a single cycle BWR steam is generated directly in the core of the reactor
at 1000 psia. A steam separator at the top of the reactor vessel directs the steam to the
turbines and the water is recirculated through the core by recirculation pumps. "Orifices at the
base of each fuel assembly control the coolant flow to the individual fuel assemblies". The
water serves as coolant and moderator. The dual cycle system also produces steam in the
core, but rather than being channeled to the turbine it goes to a steam generator where its heat
is used to produce steam in the secondary loop which is then used to turn the turbines. All
BWRs in the US have been designed and manufactured by General Electric. The predominant
single cycle type is favored because it "allows a lower capital investment for piping, heat
exchangers, pumps etc. It also has a somewhat better thermal efficiency because of the direct
generation of steam at the maximum cycle temperature". The BWR also has comparatively low
operating costs. Two problems have been identified with the BWR design used in the US, the
General Electric designed Mark 1 containment was deemed faulty and a modification was
added to prevent breech of containment in the event of a buildup of pressure within it. A
manually operated vent was added with a 300 foot stack which can be activated by operators
to release pressure from the containment directly to the environment. This prevents the
complete loss of containment by intentionally exposing the public to potentially harmful
radiation. The other difficulty is with cracking of the welds in the core shroud which surrounds
the fuel core in the reactor. This is made of circumferentially welded stainless steel plates. It
has been determined that the rate of failure in these welds has been abnormally high. When
they fail the fuel bundles are allowed to shift which can affect the ability of the control rods to
operate normally.
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Pressurized Water Reactors
Pressurized Water Reactors are the most common type of reactor used in the United States,
there are 73 operational PWRs at this time. Over half of these were designed and produced by
Westinghouse, the rest by either Combustion Engineering or Babcock and Wilcox. The
fundamental principals governing the PWR are similar to the dual cycle BWR with the
exception that the high pressure, 2250psi, in the primary loop prevents steam formation
despite temperatures of 600°F Steam is produced in the secondary loop within the steam
generators. Steam pressure in the secondary loop is lower, 1000psi. The steam is directed to
the turbines then collects in the condenser where it becomes water again and is pumped back
to the steam generator by the feed pump. The water in the primary loop acts as both coolant
and moderator. The dual cycle system means that only the water in the primary loop is
radioactively contaminated, the secondary loop as well as the turbines stay clean so radiation

5. precautions are not required near the turbines. The initial capital investment in a PWR is higher
than a BWR but operating costs are lower. Workers are exposed to less radiation due to the
duel loop design. A common problem associated with PWRs , leaks in steam generator tubes
can be caused by corrosion and high pressure. This cracking is a safety issue since a multiple
tube rupture "could result in a rapid loss of coolant accident in the reactor beyond the ability of
the Emergency Core Cooling System to control". Trojan Nuclear Power Plant in Rainier,
Oregon closed prematurely due to repeated leaks in steam generator tubes. An additional
problem is reactor pressure vessel embrittlement caused by neutron bombardment. The metal
of the reactor vessel undergoes changes as a result of the radioactivity which cause it to lose
ductility, the metals ability to expand and contract in response to variations in temperature and
pressure is lost. It is possible for an embrittled pressure vessel to lose integrity in the event of
the activation of the Emergency Core Cooling System, like pouring cold water into a hot glass
dish. Notices to plant operators on both of these issues have been sent by the NRC.
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Fast Breeder Reactor
Natural uranium consists of 99.3% Uranium-238 and 0.7% Uranium-235. Of these two forms
only Uranium-235 can be used as a nuclear fuel. In a conventional thermal reactor, during
operation some Uranium-238 is transformed into Plutonium-239 which can also be used as a
nuclear fuel. By recycling this plutonium to make new fuel it may be possible to exploit at most
about 2% of the potential fuel value of the world's natural uranium resources.
However, a fast breeder reactor can convert Uranium-238 into Plutonium-239 at a rate faster
than it consumes its fuel. By repeated recycling of the fuel it should be realistically possible to
exploit 50% of the fuel value of the uranium feed. This means that fast reactors could extend
the energy output from the world's uranium fuel reserves 25 fold.
About the term "Fast"
A neutron released from the fission of Uranium-235 (or Plutonium-239) has a high energy. To
increase the probability of the neutron causing the fission of another nucleus of the fuel
material - and thereby continuing the chain reaction - either its energy must be reduced, or the
concentration of fissionable target nuclei must be increased.
The approach of reducing the neutron energy led to the development of the nuclear reactors
which are now widely used for power production. In these reactors the energy of the neutron is
reduced by "bouncing" it off the atoms in a so-called moderator material until the neutron is in
thermal equilibrium with the atoms with which it is interacting. The neutron is then termed a
"thermal neutron" and reactors using this principle are called "thermal reactors".
The alternative approach of increasing the concentration of fissionable material and using
fission caused by high energy or "fast" neutrons led to the development of "fast" reactors.
About the term "Breeder"
If a neutron is captured by a Uranium-238 nucleus the following reaction takes place:

6. The result is that Uranium-238, which is very difficult to fission, is transformed into Plutonium-
239 which can be fissioned much more easily. This means that a useful reactor fuel can be
made from an otherwise useless natural resource. The symbol +n represents the gain of a
neutron by capture and β-
 represents radioactive decay by beta emission with the half-life
shown below the arrow.
As mentioned above, for the chain reaction to continue, on average one neutron released
during fission must go on to cause the fission of another nucleus. The average number of
neutrons released by fission depends on several factors but is usually around 2.5. Some of
these neutrons are inevitably lost as they are captured by the reactor structure or coolant but if,
on average, more than one of the available 1.5 neutrons is captured by Uranium-238 then the
total number of atoms in the reactor that can be fissioned will increase as the reactor operates.
This is the principle of the "breeder" reactor.
All reactors contain Uranium-238 in their core and so the reaction which produces Plutonium-
239 occurs to some extent in all reactors. However, only in a reactor using fast neutrons for
fission is it possible to "breed" more new fuel than the reactor consumes; this is what a Fast
Breeder Reactor does.
The amount of breeding that takes place in a fast reactor depends on the size and design of
the reactor core and the concentration of the fissile material that is in it. Some fast reactors
may be deliberately designed not to breed at all. It is important to realise that not all
fast reactors are breeder reactors.
Breeding and burning
In the early days of research in fast reactor technology it was considered very important to
maximise the amount of plutonium that was bred in order to fuel an increasing number of
reactors that would follow. Two important changes have taken place since then which have
changed this ...
Around 1960 it became clear that the economics of operating a fast reactor depend more on
minimising the number of times that its fuel must be recycled than on simply maximising the
breeding. Fuel assemblies cannot stay in the fast reactor core indefinitely for a number of
reasons, principally because the fissile material is gradually consumed. Breeding does take
place in the fuel but because it also takes place in the blanket there is a net loss of fuel in the
core centre and a net gain around the edge where it cannot contribute to powering the reactor.
In addition the steel fuel pins which hold the fuel are weakened by the gradual swelling of the
fuel pellets, chemical attack by the fuel and damage by fast neutrons. So from the early 1960's
until the end of the 1980's the emphasis was on maximising the amount of energy which could
be extracted from a fuel assembly before it had to be removed for reprocessing - the energy
extracted is usually referred to as the "burn-up" of the fuel. The results of the years of
development carried out in this field were impressive; a twenty-fold increase in the burn-up was
achieved by using oxide fuel in specially developed alloy fuel pins.

7. Since the early 1990's the picture has again changed. The widespread use of thermal reactors
means that there is now an ample stockpile of plutonium. Plutonium from dismantled nuclear
weapons may also become available with the end of the Cold War. A fast reactor which is
configured to consume plutonium is called a "burner" reactor. The fast neutrons in the core of a
fast reactor have another potential use: for a long time the disposal of a certain type of nuclear
waste produced in thermal reactors which belong to the category of elements called actinides
has been a problem. By putting this nuclear waste into the fast reactor it can be broken down
into materials which are more easily disposed of. The plutonium burning and actinide disposal
project is led by France with the participation of the United Kingdom and Japan.
The fast reactor is very versatile, a reactor can be designed so that by changing the core it can
be used to create more plutonium than it destroys or destroy more plutonium than it creates
according to what is required.
The Fuel:
As explained above, fuel cannot stay in the reactor indefinitely. To recover the useful materials
in the fuel it must be sent to a reprocessing plant for chemical separation. The recovered
unburnt and bred plutonium is then returned to the fuel fabrication plant to be put into new fuel
rods.
Fast Reactor Design
The coolant
The coolant which passes through the core of the fast reactor must not moderate (slow down)
the neutrons emitted from the fission reaction. This limits the choice of coolant slightly, in
particular water is unsuitable as a coolant because it is a very effective moderator. To avoid
moderating the neutrons suitable coolants are liquid metals and inert gases (e.g. helium). Little
research has been done on inert gases in fast reactor applications, liquid metals are the
preferred option due to their excellent heat transfer properties.
The liquid metal coolant would preferably remain in the liquid state below or as close as
possible to ambient temperatures (to avoid expensive pre-heating of coolant filled vessels and
pipes), have a boiling point which gives a margin of safety above the proposed reactor
operating temperatures and be obtainable in sufficiently large quantities at a reasonable cost.
Possible choices of liquid metal coolant would include mercury, lead, sodium and a sodium-
potassium mixture; except for lead, all these have indeed been used as fast reactor coolants
and designs with lead coolant have been made. However, mercury and lead have problems of
chemical toxicity which make them unattractive and mercury is also very expensive. The
sodium-potassium mixture - often referred to as NaK - has the advantage of remaining liquid at
room temperature, obviating the need for pre-heating, but in other respects is more difficult to
handle than sodium. Fast reactor engineers around the world have therefore arrived upon the
same conclusion that the best coolant is sodium.

8. The reactor layout
One of the most important choices facing the designer of a fast reactor is the layout to be
adopted for the primary circuit which includes the reactor. There are two main concepts, the
"pool" and the "loop". In a pool-type reactor one large vessel holds the core, the Intermediate
Heat Exchanger (IHX) which passes heat to the secondary loops, and the pump which
circulates the primary sodium. In a loop-type reactor, the core, IHX and pump are each in their
own smaller vessels linked by pipes. The layouts are illustrated below, sodium flows upwards
through the core (shown yellow) in both designs:
The choice is not simple since each has its own advantages and disadvantages. The vessel of
the pool is a very simple design with no branches to cause stress concentrations. It can be
arranged so that hot coolant never comes into contact with the vessel wall. The disadvantages
of the pool are that the vessel is so large that it must be fabricated on-site where quality
assurance is more difficult. Once in operation its internal structures are difficult to inspect. The
reactor vessel of the loop-type, being much smaller, can be built in a factory and transported to
the site. The pipework of the loop reactor may be longer and more complicated but it is easier
to inspect.
Both types of reactor have been built but, to date, there is less experience with large scale
loop-type reactors. MONJU which is a loop-type reactor will provide an interesting comparison
with the European pool-type prototypes.
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