future of fast breeder reactors final

8/2/2019 Future of Fast Breeder Reactors Final

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UNIVERSITY OF TEXAS AT EL PASO

Breeding Energy For The

Future

Fast Breeder ReactorsLuis Iturralde 800417211

3/22/2010

An energy crisis is affecting the world; it is not conceivable to keep relying only on fossil fuels for

energy, either because of limited availability or the need to protect the environment. Renewable energy

sources are necessary to help attain and secure energy resources for the future. The development of new

technology can allow us to produce safer and more efficient reactors that have the potential of solving the

energy crisis.


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Outline

Breeding Energy for the Future

1. Preface2. Introduction3. Nuclear Power Background

3.1.How it works3.2.History3.3.Availability

4. Types of reactors4.1.Light-Water Reactors (LWR)

4.1.1. Pressurized Water Reactors (PWR)4.1.2. Boiling Water Reactors (BWR)

4.2.Pressurized Heavy Water Reactors (PHWR or CANDU)4.3.High Power Channel Reactor (RBMK)4.4.Gas Cooled Reactor(GCR) and Advanced Gas Cooled Reactor (AGR)4.5.Liquid Metal Fast Breeder Reactor (LMFBR)

5. Fast Breeder Reactors5.1.How FBRs work

5.1.1. Types of breeder reactors5.1.2. Availability5.1.3. Concerns5.1.4. Future

5.2.New technology on breeder reactors6. Conclusions7. References


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Future of Fast Breeder Reactors

2. IntroductionFast breeder reactors are a sustainable source of energy for the world because they have

the ability of producing more of its own fuel. Producing energy along with breeding fuel is the

main objective or reason for developing fast breeder reactors so a long-term fuel supply can be

achieved. Breeding its own fuel allows a breeder reactor to become a sustainable source of

energy. Reducing the actinides in the nuclear waste is another reason for the research of fast

reactors along with the interest of taking advantages of their high thermal efficiency. Breeders

do not differ much from traditional reactors, their main difference is that they do not use a

moderator; the lack of a moderator is what enables the reactor to produce or breed more

fissionable material. FBRs have the potential of producing less or even eliminate waste by

burning it again in a closed cycle (Newman). There are many concerns about nuclear energy in

general, but there is also a lot of opposition to Fast Breeder Reactors, because they are said to be

unstable and they can be used to breed plutonium for use on nuclear weapons.

Taking a look at nuclear power we can find that it is a very viable solution or alternative

to the world dependency on oil. It is possible to produce power from controlled nuclear reactions,

nuclear plants that produce electrical energy use the power from fissile reactions to heat water

and produce steam which is then used to operate turbines which in turn produce electricity. 15%

of the worlds electricity comes from nuclear power as of 2009, and more than 150 naval vessels

that operate on nuclear power have been built. (Wiki) There are many concerns about safety and

radioactive waste management, but new technology will allow us to increase safety measures,

and elaborate better ways of managing the radioactive waste. The use of breeder reactors would

guarantee the availability of fuel for many years.

Enriched uranium or another fissile material must be used in starting a fast breeder

reactor. A breeder reactor differs from a regular reactor because it produces, or breeds, more

fuel than it consumes. On average a breeder reactor may produce 500lbs of plutonium. A breeder

reactor can be used like any other reactor to produce electricity. The first fast breeder reactor

was built in the US in 1951, it produced a minimal amount of energy about 0.2 MWe and it was

in operation for about 12 years. It was succeeded by a two 20 MWe, a 60 MW and Fast Flux TF

which had a thermal output of 400 MW from 1980-1993. A 15 MWe fast breeder reactor

operated from 1959-1977 in the UK followed by a 270 MW that operated from 1974-1994.


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France built its first Fast Breeder Reactor in 1966 which had an output of about 40 MW thermal.

Then the Phenix was built in 1973 with an output of 250 MWe and is still in operation. The

French also built the Superphenix in 1985 and operated until 1998, with an output of 1240 MWe.

It is the biggest reactor of the world but was closed due to safety concerns. Germany had a very

small breeder reactor with an output of only 21 MWe and operated from 1977 to 1991. India has

one that produces 40MW thermal and was built in 1985. The Joyu built in Japan in 1978 has an

output of 140 MW, Japan also had the Monju which was closed in the year 1996 due to safety

concerns and operated for only 2 years with an output of 280MWe. (Newman)

3. Nuclear Power Background

In 1895 Wilhelm Rontgen discovered ionizing radiation when he passed an electric

current through an evacuated glass tube. Henri Becquerel found that an ore containing radium

and uranium caused a photographic plate to darken, later he demonstrated that beta radiation and

alpha particles that where being emitted caused this effect. Gamma rays, another type of

radiation, were found by Villard. The name radioactivity was given by Pierre and Marie Curie

to this phenomenon, all this happened during the year of 1896. Ernest Rutherford in 1902

demonstrated that radioactivity could create a different element by emitting an alpha or beta

particle form the nucleus. Rutherford developed a fuller understanding of atoms until in 1919 he

found that nuclear rearrangement occurred when he bombarded nitrogen with alpha particles

from radium and oxygen was formed. The understanding of the atom and how the electrons are

arranged around the nucleus was increased by Niels Bohr through the 1940s. (World Nuclear

Association)

Naturally-radioactive elements have a number of different isotopes, with the same

chemistry, this was discovered by Frederick Soddy in 1911; same year in which those isotopes,

also called radionuclides, were found by George de Hevesy to be invaluable as tracers, because

minimal amounts can be easily detected with simple instruments. The neutron is discovered by

James Chadwick in 1932 during the same time Cockcroft and Walton where producing nuclear

transformation by shooting accelerated protons into atoms. Irene Curie and Frederic Joliot found

in 1934 that Cockcroft and Walton transformations created artificial radionuclides. A much

greater variety of artificial radionuclides could be formed when neutrons were used instead of

protons, Enrico Fermi discovered the following year. Fermis experiments continued, and he was


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able of producing heavier elements, but when using uranium he is also able of producing lighter

ones. By the end of 1938 it was shown that those new lighter elements were barium and other

about half the mass of uranium, thereby demonstrating that atomic fission had occurred, this was

discovered by Otto Hahn and Fritz Strassman. This was explained by Lise Meitner and Otto

Frisch by suggesting that the neutron was captured by the nucleus, resulting in severe vibrations

which lead to the splitting of the nucleus into two not identical pieces. Energy release from

fission was calculated to be of about 200 million electron volts. This figure was later confirmed

by Frisch experimentally at the beginning of the year 1939. This was the first experimental

confirmation of Albert Einsteins paper putting forward the equivalence between mass and

energy, which had been published in 1905. (World Nuclear Association)

All the discoveries and developments that occurred during 1939 motivated many

laboratories, Hahn and Strassman showed that fission in addition to releasing lots of energy also

releases additional neutrons that can cause fission in other uranium nuclei and the possibility of

sustaining a chain reaction. Joliot in Paris and Leo Szilard working with Fermi in New York

confirmed this suggestion. It was soon proposed by Bohr that fission was much more likely to

occur in the uranium-235 isotope than in U-238 and that slow-moving neutrons are more

effective in producing fission. Szilard and Fermi confirmed that slow neutrons are better for

fission and proposed using a moderator which would slow down the emitted neutrons. These

ideas were extended by Bohr and Wheeler into what became the classical analysis of the fission

process, they were able to publish their paper just two days before the beginning of war in 1939.

U-235 was known at that time to consist of only 0.7% of natural uranium, the rest being U-238.

So it was considered that the separation of the two would be difficult. This increase in the

proportion of the -235 isotope became known as enrichment. In the time period from 1939-

1945 almost all the development was focused on producing the atomic bomb. During 1945 the

attention was turned into developing a way of controlling this energy for producing electricity

and naval propulsion. By the end of World War II, development of nuclear weapons continued,

but a new focus emerged, harnessing the amazing nuclear power for making electricity from

steam. During the race for developing nuclear weapons the world had acquired many new

technologies, and scientists realized that the tremendous heat that released during the process

could be seized to produce electricity. It was also thought that the new form of energy could be

developed into a small size and could have many applications, mainly in ships and submarines.


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The small Experimental Breeder reactor (EBR-1) in 1951 was the first nuclear reactor to produce

electricity, a minimal amount, in Idaho. Atoms for Peace was proposed by President

Eisenhower in 1953. This program focused on changing the direction of the research efforts

toward the generation of electricity, and it gave course to a civil nuclear development in the USA.

On the other side of the world, the Soviet Union started modifying its reactors. Their existing

reactor that used graphite as a moderator was modified for heat and electricity generation instead

of serving its original purpose which was to produce plutonium for nuclear weapons. The

worlds first nuclear powered electricity generator started up in 1954. It was a water cooled

reactor, moderated by graphite, was called the AM-1 which translated means peaceful atom. It

had an output of 30 MW thermal or 5 MW electric. Very similar in to the plutonium producing

reactors used for military purposes, and served as a prototype for other graphite channel reactors,

including the Chernobyl-type high power channel reactor, or RBMK (reaktor bolshoi

moshchnosty kanalny). The AM-1 was used until 2000 as a research facility but only produced

electricity for about 5 years. BR-1 (bystry reaktor) a fast neutron reactor began functioning in

1955, but it never produced power but its development conduced to the construction of the BR-5.

This new breeder reactor had a capacity of 5MWt; it started up in 1959 but was only used for

research necessary for the design of sodium-cooled FBRs. (World Nuclear Association)

Admiral Hyman Rickover, led the US main effort, he developed the Pressurized Water

Reactor for naval use, mainly in submarines. Enriched uranium was the fuel used in the PWR

and it was moderated by ordinary water. In Idaho, during March 1953, the Mark 1 prototype

naval reactor was started up. Later in 1954 the first nuclear powered submarine was launched,

the USS Nautilus. During 1954 the US Atomic Energy Commission built in Pennsylvania the

60MWe Shippingport demonstration PWR reactor which operated until 1982. Development in

the UK took a different approach which resulted in various reactors, Magnox type reactors,

fuelled by uranium metal, moderated by graphite and cooled by gas. Calder Hall-1 started

operating in 1956 and worked until 2003 with an output of 50MWe, this was the first of the

series of Magnox reactors created in Britain since the USA had a monopoly on enriched uranium.

They reached a maximum of 26 units and after q963 no more were built. Next the British

focused on the Advanced Gas cooled reactor and later finally realized the practical virtues of the

PWR. (World Nuclear Association)


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Around 1960 nuclear energy becomes commercial, Westinghouse developed the first

commercial pressurized water reactor. Yankee Rowe started operating, with an output of

250MWe and ran until 1992. In the meantime Argonne National Laboratory was developing the

boiling water reactor (BWR), Dresden-1, the first of its kind, had an output of 250MWe, was

designed by General Electric and started operation in 1960. Many orders for PWRs and BWRs

with outputs of about 1000MWe were placed by the end of the decade of the 60s. The first

Canadian reactor was a CANDU design, which utilizes natural uranium as fuel and heavy water

as moderator and also as a coolant, the first one was created on 1962. The French started with a

design similar to that of a Magnox and their first reactor started up in 1956. The Soviet Union in

1964 commissioned its first two nuclear power plants. With an output of 1000MW the first large

RBMK reactor was started by the soviets in 1973. The worlds first commercial prototype fast

neutron reactor, the BN-350 was built in Kazakhstan and started up in 1972 producing 120MW

of electricity and heat that is used to desalinate water from the Caspian Sea. Some countries like

the UK, USA, France and Russia

had a number of experimental fast

breeder reactors that produced

electricity since 1959, but by 2009

all of them have been shut down.

Many countries around the world

have preferred light water reactors

to use on their respective nuclear

programs. Today 60% of the

worlds reactors are Pressurized

Water Reactors and 21% are

Boiling Water Reactors. Nuclear

power development has suffered a decline since the high 1970s to about 2002. During this

period the creation of new reactors barely surpassed the number of reactor that where being

retired from service. Nevertheless the improvements made in technology caused the capacity to

increasing by nearly one third and the output increased by more than a half. The increased

demand for electricity worldwide, being aware of the importance of securing energy and the

concerns about global warming are the three of the reasons why the nuclear power is recovering.

Figure 1 (wiki commons)


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The availability of a new generation of reactors also contributes to this Nuclear Renaissance.

In Europe, specifically in France, there are plans to build new reactors that will replace what they

have available now, also there is a 1600 MWe PWR in order for Finland, similar to this one are

those planned for France. In 2005 the Energy Policy Act gave incentives for establishing new-

generation power reactors. Plans in Asian countries like China, India and Japan, are much more

adventurous of those of Europe and North America. China alone has an overwhelmingly larger

number of reactors planned for 2020 (World Nuclear Association).

4. Types of Reactors

Several different types of reactors exist; this can be classified in various ways. In this

portion the main kinds of reactors will be described in order to provide a simple understanding of

how they function and their main differences, it is also necessary to discuss the advantages and

disadvantages of each one of the different designs.

4.1 Light-Water Reactors (LWR)

This category of reactors can be separated into two main subcategories the Pressurized

Water Reactors (PWR) and the Boiling Water Reactors. These two are called light water reactors

because the operate using regular water as their moderator and coolant.

Pressurized Water Reactors

This kind of reactor is a thermal neutron reactor, meaning that the neutrons are slowed

down by a moderator to ensure or increase

the probability that fission occurs. The

reactor is made up of a pressure vessel that

contains the fuel, control rods, and coolant.

High pressure liquid water is used as the

coolant and moderator of choice in this

reactor. This design is composed of two

circuits of water flowing throughout the

reactor. The primary cooling circuit flows

through the core of the reactor at very high Figure 2 Diablo Canyon PWR (Wikipedia)


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pressure, approximately 16MPa, this pressure prevents water from boiling inside the reactor; and

the secondary circuit in which is used to generate steam. The steam then drives a turbine which

in turn drives a generator to produce electricity. The use of water in PWRs is an important safety

feature, because as the temperature increases water becomes less dense hence reducing the extent

to which the neutrons are slowed down and thereby reducing the reactivity of the reactor, this

feature makes PWRs very stable (source). The fuel used in PWR is enriched uranium dioxide

(UO), usually in the form of hard ceramic pellets. These cylindrical pellets are then arranged in

bundles which are used to build the reactor core. Typically a reactor has assemblies of about

200-300 rods each and are 4 meters in length; and a large reactor uses between 150-250

assemblies, totally in about 80-100 tons of uranium.

Boiling Water Reactor (BWR)

This is the second kind of light

water reactors. This design is very similar

to the PWR, but this reactor has only one

water circuit and water is at a lower

pressure, this reactor can operate, in fact is

designed to work with up to 15% of the

water as steam in the top part of the core,this gives it less moderate and as a result

less efficiency. The water vapor is passed

through steam separators and then it goes

directly into the turbine, which because of this is part of the reactor circuit. The turbine needs to

be shielded because all the water that is in contact with the reacts has traces of radioactive

contaminants; this extra cost of shielding the turbine is balanced with the simpler design of the

reactor. Most of the radioactivity in the water is very short-lived, most of it being N-16, with a 7

second half-life, so the turbine hall can be entered soon after the reactor is shut down ( World

Nuclear Association).

4.2 Pressurized Heavy Water Reactor (PHWR) or CANDU (Canada Deuterium

Uranium)

Figure 3 Laguna Verde Nuclear Power Plant (BWR) (Wikipedia)


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Canada has been developing

the PHWR since the 1950s reason

because it is also known as the

CANDU; recently India has started

working on it. In the CANDU

reactor the moderator is enriched

instead of the fuel. Natural uranium

oxide is used as fuel in these

reactors, therefore it needs a more

capable moderator, reason why heavy water is used (DO). The moderator is stored in a tank

called calandria which is penetrated by several hundred horizontal pressure tubes which

contain the fuel. PHWRs can be refueled while functioning at full power.

4.3 RBMK (High Power Channel Reactor)

These reactors were designed by the

Soviet Union and have the capabilities of

producing both power and plutonium. Water is

used as a coolant and graphite as the moderator,

and is fueled by un-enriched uranium oxide. Fuelis assembled into 3.5 meter long rods. It lets water

boil much like in a BWR, also employs a pressure

tube design different from the pressure vessel

used in the PWR. According to the International

Nuclear Safety program the biggest strengths of the RBMK are the low core power density that

they have provides a unique ability to withstand station blackout and loss of power events of up

to an hour with no expected core damage and they can be refueled while operating like the

CANDU also the use of graphite as a moderator allows the use of a form of uranium that is not

suitable for use in the light water reactors. RBMK are large and unstable this makes them

expensive and dangerous (World Nuclear Association). Some safety flaws have been identified

and corrected as a result of the Chernobyl disaster. As of this year only 11 RBMK continue to

operate in the world (http://insp.pnl.gov).

Figure 4 CANDU Qinshan Nuclear Power Plant (Wikipedia)

Figure 5 Smolensk RBMK (www.industcards.com)


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4.4 Gas Cooled Reactor (GCR) and Advanced Gas Cooled Reactor (AGR)

Thermal neutron reactors, that operate using carbon dioxide (CO) as coolant and the

moderator used is graphite. GCRs and AGRs are capable of operating at higher temperatures

than PWRs giving them higher thermal efficiency as stated by the World Nuclear Association.

Uranium oxide pellets in stainless steel tubes are used as fuel for an AGR which are the second

generation of British gas-cooled reactors. Carbon dioxide at 650C circulates trough the core and

also trough the steam generator tubes outside it,

all of which are still inside the pressure vessel

made of concrete and steel. There is a secondary

shutdown system that involves injection nitrogen

into the coolant. Many of these reactors operate

in the UK where the concept was developed based

on the Magnox reactors and there are also few of

the original Magnox reactors, but all will be shut

down by 2010

(www.no2nuclearpower.org.uk/reports/agrs.php).

4.5 Liquid Metal Fast Breeder Reactor (LMFBR)

Last kind of reactor to be presented and the

main focus of this report, is the Fast Breeder

Reactor. This kind of reactor is a classified as a fast

neutron reactor, since they do not employ any

moderator to slow down neutrons. The most

common type of breeder reactor is the Liquid Metal

Fast Breeder Reactor (LMFBR) which is cooled by

liquid metal. The main characteristic that separates

breeder reactors from the rest is the fact that they

produce more fuel than what they consume in other

words they breed. FBR operate using the U-238

isotope which is the most common isotope of uranium in the world, even tough, they get 60

Figure 6 Dungeness 2 an AGR

(www.no2nuclearpower.org.uk/reports/agrs.php).

Figure 7 Superphenix FBR in France

(www.elrst.com/wp-

content/uploads/2009/11/Superph%C3%A9nix-

fast-breeder-france.jpg)


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times as much energy compared to the typical reactor, fast neutron reactors are expensive to

build and the high cost of reprocessing fuel makes FBRs economically impractical, because

prices of uranium of over 200 us dollars would be required in order to make them competitive

with regular reactors according to the World Nuclear Association. It has been estimated that

there is enough uranium (U-238) available to operate this nuclear plants for up to five billion

years (McCarthy)

5. Fast Breeder Reactors.

The World Nuclear Organization states that Fast neutron reactors are a technological

step beyond conventional reactors. As mentioned before breeder reactors receive that name

because they can produce more fuel than what they consume. These are of the fast neutron

design, since they do not require a moderator and are usually cooled using liquid sodium or lead.

Fast Breeder Reactors offer a more efficient use of uranium resources they also have the ability

of burning actinides which are the long-lived components of high- level nuclear wastes. FBRs

original purpose was to burn uranium more efficiently and consequently extends the worlds

resources of the element. Geological exploration by the 1970s showed that uranium resources

were not as scarce as it was assumed before, some many of the countries that had done extensive

research decided to focus on another kind of reactor, which had fewer problems and was cheaper

to build. Technical advancement has been made, but the economics is the main reason why the

development of FBRs has clogged. Next the working features and details of a Fast Breeder

reactor will be discussed.

5.1 How FBRs work

In a fast reactor fission is the solely responsibility of the fast neutrons, that is why the

most used fuel is plutonium, since fast neutrons are not as efficient as slow neutron to create

fission in uranium, but they are capable in plutonium. Conventional fast breeder reactors have a

fertile blanket of depleted uranium (U-238) around the core, where much of the Pu-239 is

produced. The plutonium produced in the core remains pure Pu-239, the blanket can be

reprocessed and the plutonium recovered (Fast Neutron Reactors WNA). The core in most FBRs

is actually made up of two different parts, the core itself and a blanket, in which most of the

breeding occurs, the core of a fast breeder reactor is much more compact than the core of a


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PWR or a BWR. Fission takes place in the core but the extra neutrons are absorbed by the

blanket surrounding the core. As an example is possible to assume that 100 fissions produce 300

fast neutrons, then the results will be as follows; 100 neutrons keep on going the chain reaction,

another 100 breed or convert U-238 into Pu-238 in the core, of the 100 neutrons that are left,

40 are lost by parasitic absorption, the remaining 60 leak through core into the blanket and 50

are converted into plutonium the rest are lost again because of parasitic absorption. The breeding

ratio in the core is only of about 0.8 while in the blanket can be around 1.25 (power point). Rods

filled with uniform material compose the fuel assemblies in the blanket while the fuel assemblies

in the core are made up or rods whose central segments are packed with fissile material, while

the end sections contain fertile material. Fuel assemblies from the core and blanket are removed

from the reactor for reprocessing; plutonium is detached from them and can be used as fuel.

Because of high heat generation a coolant with excellent heat transfer properties is required,

reason why the common choice are liquid metals or even pressurized helium, some typical

choices are sodium, lead, some lead and sodium mixtures and even mercury, liquid metals give

better heat transfer but pressurized helium does not slow down neutrons as much. The size of the

reactor can influence the coolant choice, because a small reactor cores require high fuel density

which is better achieved by using liquid metals as coolant in the reduced space; while larger fast

breeder reactors, like for commercial use power plants, do not require such high density fuel and

the space available also allows for the use of pressurized helium. The best choice, or at least the

common one, is sodium because of its high specific heat (

), high boiling point (883)

low pumping power requirement, low system pressure requirements and its ability to absorb

considerably energy under emergency conditions among other; however sodium reacts violently

with air and water, activates under irradiation and has neutron decelerating (slight) and

absorption properties (Ref. Khodarev). For LMFBRs there are two basic types, the pool or

integrated type and the loop type. The pool design is very simple, the vessel contains not only the

core but also some other components, almost all the components, the core, cooling pumps and

heat exchanger, are submerged inside a liquid metal (sodium) pool at approximately atmospheric

pressure, doing it this way reduces the amount of piping required. One disadvantage of this is

that the pool is very large and while operating the inspection of the structure is complicated since

they operate submerged in liquid sodium or another liquid metal. The loop type looks similar to a

light-water reactor, because all of the cooling system components are outside the vessel, which is


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smaller and only contains the core and core blanket, but the piping for it is more complicated,

extensive and expensive but inspection are not nearly as complicated as with the pool type. Both

arrangements guard the vessel and primary components with a guard vessel around it so in case

of a rupture of any of the primary components it does not lead to a huge spill of liquid

radioactive sodium. One of the main targets in any fast breeder reactor design is to decrease the

stoppage time necessary for refueling.

Figure 8 Schematic of LMFBR (en.wikipedia.org/wiki/File:LMFBR_schematics2.svg)

A rotating plug is commonly used in both types of reactor design; the plug is located at the top of

the vessel in the closure head. An in-vessel fuel transfer machine and control rods are mounted

on the rotating plug, control rods are disconnected from the core before the plug is rotated; this

makes it possible to transfer fuel from the core to any point inside the reactor and the other way

around. A temporary storage drum is located inside the reactor vessel in the pool design, where

the spent fuel is commonly placed and it remains there until the decay heat is removed. Thevessel fuel machine can later be used to transfer spent fuel to storage outside the reactor and this

process can be done while the reactor is operating. Spent fuel is directly removed from the core

to an external storage facility in the loop type reactor. (Ref. Khodarev).

Availability
http://upload.wikimedia.org/wikipedia/commons/4/46/LMFBR_schematics2.svg


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For about 50 years fast breeder reactors have been under development, there are

important programs for the development of fast breeder reactors in several countries. With a

thermal efficiency of 43-45% two fast breeder reactors are being operated currently, being this

the highest efficiency values in the nuclear power industry. Effective breeding rations and the

closed fuel cycle has been demonstrated experimentally. Is said by the International Atomic

Energy Agency that in total FBRs have accumulated 300 reactor years of operation, that at

several reactors fuel burn up in excess of 130,000 MWd/t has been reached and commercial

reactor designs have been greatly advanced. The understanding of liquid metal fast breeder

reactors safety requirements or needs has dramatically increased during the last few decades

trough the extensive research that has been made. Safety research from the past has been used

efficiently in developing methods of safety analysis which were used to evaluate the safety of

present and new and advanced reactors. It is believed that a high degree of safety can be reached

by the liquid metal cooled fast breeder reactors that at this time are being planned. A big step

toward the full commercial utilization of fast breeder reactors, generally consistent with other

studies indicates that competitive reactors are not far away (LMFBR IAEA).

Liquid metal reactors in France help to demonstrate positive cases of designs, realized

projects and experience in FBRs construction and operation. The Rapsodie, an experimental fast

breeder reactor which operated from 1967 to 1983 and had an output of 40 MWth, the Phnix

reactor whit an output of 255 MWe that started operations in 1973, and the Super-Phnix (1986-

1998) are examples of this positive situation in the history of using fast breeder reactors.

Reprocessed fuel from FBRs in France is about 30 tonnes cumulative. For the Phnix a breeding

ration of 1.16 mas confirmed; it was operated at a temperature of for 100,000 hours and

had a thermal efficiency of 45.3%; burn up was increased to a maximum exceeding

150,000. All these levels were reached using 166,000 fuel pins to conform eight cores of

fuel. The Phnix operations were resumed on 2003 after a plant renovation program was

completed and as of that year its power was limited to , . The last fuel

assembly of the Super-Phnix reactor was unloaded on March 2003; its shut down process

started in 1999 (LMFBR IAEA).

Event tough the development of liquid metal fast breeder reactors in Europe has been delayed

alternative applications of fast reactors, mainly the transmutation of long-lived nuclear waste and

the utilization of the extra plutonium, are being developed in many countries like; France,


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Germany and the UK. On April 1980 the BN-600 fast breeder reactor was connected to the grid,

reaching its full power by October 1981 its operations are stable with a turbine efficiency of

about 43 percent. The Russian BN-600 has generated over 91 billion kWh of electricity by the

end of 2004. Currently Russia has focused its effort into increasing the safety and improving

economics in the reactor operations. The design was completed of a commercial size fast reactor,

the BN-800, and the construction license has been issued. The startup of the BN-800 reactor is

scheduled for 2010 at the BN-600 reactor site in Beloyarskaya. BN-1800 is the next big step for

Russia in the development of fast breeder reactors. (LMFBR IAEA)

Currently the UK government has no program for fast breeder reactors design and development

but there is a plan privately funded by some companies like BNFL. Specific and key areas are

the focus of the UK investigation, like: nuclear methodology and core design, fuel performance

and fuel cycle modeling (LMFBR IAEA).

Th Argonne National Laboratory has developed the sodium-cooled integral fast reactor (IFR) in

the USA. This project focuses on the use of fuel composed made by alloying U-Pu=Zr for

loading its core. General Electric integrated the IFR into a full plant design of a 300 MWe

advanced liquid metal cooled reactor (ALMR) in which the plutonium is not separated from

higher actinides they are recycled together in the reactor and never leave the reactor site

(LMFBR IAEA).

Waste

The International Atomic Energy Agency has identified and analyzed some of the

concerns with fast breeder reactors. One of the main concerns for any nuclear power plant is the

waste management. There is lots of preoccupation about pollution generated from nuclear waste.

Among the solutions for nuclear waste management is the implementation of new fast breeder

reactors which will burn all the long-live actinides, there is also lots of emphasis and new waste

disposal and management methods and no leaving out the reprocessing technologies that are

being developed and expected to be deployed in conjunction with newer neutron reactors.

Since the beginning of the use of nuclear power, the main reason for reprocessing used

fuel was to recover unused uranium and plutonium and right there ending the fuel cycle. Another

reason is to reduce the long term radioactivity of the high-level waste. This also reduces the

possibility of using plutonium for weapons. MOX fuel is mainly composed of recycled

plutonium, actually most of the recycled plutonium is used for MOX fuel, but only a small


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amount of recovered uranium is recycled. There is a growing interest in recovering all long-lived

actinides and recycle them in fast breeder reactors so they end up as short-lived fission products.

Minor actinides are not destroyed in in the recycling process through LWR, so the focus for the

future is to burn them in fast neutron reactors. So the reprocessing and recycling processes in

todays reactors are only a provisional part of nuclear power development with the general use of

FBRs pending. Five of the six new reactor technologies that are being considered include fuel

cycles that recycle all the actinides. The US policy has always avoided reprocessing, but on 2006

$50 million were designated to develop integrated spent fuel facilities and the way for

achieving it using FBRs is clear now (Processing of Used Nuclear Fuel).

The position statement release by the American Nuclear Society says that:

The American Nuclear Society believes that the development and deployment of

advanced nuclear reactors based on fast-neutron fission technology is important to

the sustainability, reliability, and security of the worlds long-term energy supply

Fast reactors in conjunction with fuel recycling can diminish the cost and

duration of storing and managing reactor waste with an offsetting increase in the

fuel cycle cost due to reprocessing and fuel refabrication. Virtually all long-lived

heavy elements are eliminated during fast reactor operation, leaving a small

amount of fission product waste that requires assured isolation from the

environment for less than 500 years. (American Nuclear Society Position

Statement)

Future

Electricity (experimental) from nuclear power using a fast reactor was first produced

more than 59 year ago, since then many fast reactors have existed most as experimental and only

a few for actual use. Since then many improvements have been made and lots of experience has

been gained on their maintenance, safety and operation. By some moments FBR development

and even investigation almost stopped completely but circumstances amongst which are the

energy crisis and the need for cleaner energy have contributed to a revival of this technology.

China is building the experimental fast reactor (CEFR), which represents the first move towards

fast breeders technology in China. The construction has been completed as of 2009 and

originally its first criticality was expected for 2008, but it was delayed and no further notices of


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when criticality will be achieved could be found. The CEFR has an output of 65 MWth/25MWe.

China has planned 20 new reactors, including some of the most advanced in the world and are

expected to be operational by 2020 and the main focus is on breeder reactors technology.

Another effort from China on fast reactor technology is the China Prototype Fast Reactor

(CPFR), which is under construction, with a main focus on the minor actinide transmutation

(Nuclear Power in China).

India has an important and flourishing nuclear power program that aims to have 20,000

MWe nuclear capacity by 2020 and finally produce 25% of electricity from nuclear power by

2050. Based on its knowledge in fast reactors and thorium fuel cycle India expects to turn into a

world leader in nuclear technology. However, due to its weapons program, India is outside the

Nuclear Non-Proliferation Treaty which has excluded them from nuclear plant or materials trade.

Currently India has 5 nuclear power reactors of which one is a Fast Breeder Reactor, the

Kalpakkam PFBR. Indias Kalpakkam is a 500 MWe prototype FBR that is underconstruction

by Bharatiya Nabhikiya Vidyut Nigam Ltd which is a government enterprise set up to focus on

FBRs. It is expected to be completed and start operations by the end of this year and start

producing electricity by 2011. The PFBR will be fuelled with uranium-plutonium oxide and will

have a blanket with thorium and uranium to breed fissile U-233 and plutonium. Thorium

program gets a boost from this design and it seems possible that eventually India will be able to

take full advantage of its large thorium reserves. Indias main objective is to develop advance

heavy-water reactor for the thorium cycle. By 2017 four more FBRs are expected to be built,

two at the same site (Kalpakkam) and two more at another site and in total there are plans for six

FBRs in India. These initial reactors are expected to be fuelled with mixed oxide or carbide, but

the following will be metallic fuelled. Of those six FBRs mentioned above one will be a dual fuel

unit, meaning that it will have the flexibility of converting from MOX to metallic fuel. A 1000

MWe fast reactor is expected to follow the previous; its construction is expected to start around

2020 and will use metallic fuel. By 2014 a fuel fabrication plant and a reprocessing plant for

metal fuels are intended for Kalpakkam. The Nuclear Desalination Demonstration Plant (NDDP)

at Kalpakkam was commissioned in 2002 to help address the shortage of water in coastal regions;

it consists of a Reverse Osmosis (RO) unit of 1.8 million liters per day (Nuclear Power in India).

Japans Joyo experimental reactor which has been operating since 1977 is will be boosted to 140

MWth. Japan plans to develop new FBRs like its Monju prototype commercial FBR which


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originally was connected in 1995 but was then shut down due to a coolant leak but there are

plans for reviving it. Japan is also building its Japan Standard Fast Reactor. This unit would

have an output anywhere between 500 to 1500 MWe and is intended to burn actinides and used

plutonium and uranium fuel. Mitsubishi Heavy Industries is in charge of building it (Nuclear

Power in Japan).

Six reactors are under construction in Russia, but only one is a Fast breeder reactor. Russia has

extensive experience on operations of fast breeder reactors, since its BN-600 has been supplying

electricity since 1981 and has the best operating and production record of all Russias nuclear

power units. It is cooled by liquid sodium and uses uranium oxide as fuel, and there are plans to

reconfigure the BN-600 to burn the plutonium from military stockpiles. For 27 years in

Kazakhstan the BN-350 FBR operated using half of its output for water desalination. A new

larger 800 MWe fast breeder reactor is being built at Beloyarsk, the BN-800. With many

improved features like fuel flexibility (can use U+Pu nitride, MOX or metal) breeding ratio of up

to 1.3 and enhanced safety and economy it is a great improvement if FBRs design. It has

capabilities of burning two tons of plutonium from dismantled weapons per year it will also help

on testing actinides recycling in the fuel. Such improvements made it the choice of reactor for

China and two of these units are expected to be built there. The BN-800 is probably the last

reactor with a fertile blanket to be built in order to minimize the risk of weapons proliferation

(Nuclear Power in Russia).

For over 40 years Russia has used lead-bismuth coolant for the reactors in its submarines.

The BREST is a new Russian design of 300 MWe or more using lead as the primary coolant. It

uses high density U+Pu nitride fuel, and it cannot produce weapons grade plutonium since it

does not have a uranium blanket because all the breeding takes place in the core. With onsite

reprocessing fuel can be recycled indefinitely. Plans include a pilot unit in Beloyarsk (Fast

Neutron Reactors).

ELSY, the European Lead-cooled System fast neutron reactor with high flexibility, it can use

depleted uranium or thorium and burn actinides from LWR fuel. Pb or Pb-Bi liquid metal at low

pressure is used for cooling. A small-scale demonstration facility is planned, it will run on MOX

fuel at 480 and the molten lead is then used in steam generators, this project is financed by

EURATOM and led by Ansaldo Nucleare from Italy (Fast Neutron Reactors).


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General Electric was designing a modular Liquid metal-cooled inherently-safe reactor (PRISM)

during the advanced liquid-metal fast breeder reactor program (ALMR). GE-Hitachi have

designed the PRISM today, it is a modular- pool-type reactor with passive cooling for decay heat

removal. The PRISM is GEHs Generation IV solution. Consisting of two modules of 311 MWe

each, the PRISM power block operates at high temperature. The complete primary system is

contained below ground level. Metal fuel is used (Pu & DU) and can be obtained from used light

water reactor fuel. One third of the fuel is removed every two years, with fuel staying inside the

reactor for a total of 6 years. The concept of a commercial-scale plant is made up to three power

blocks that provide 1866 MWe. After removing fission products the used fuel form PRISM can

be recycled (Generation IV Reactors).

Korea Advanced Liquid Metal Reactor (KALIMER) is a 600 MWe sodium cooled fast reactor of

the pool design. This is Koreas approach to generation IV reactors but it still requires future

development. The KALIMER has no breeding blanket and a transmuter core (Fast Neutron

Reactors).

For France a prototype of generation IV reactor was announced on 2006 to be operating on 2020.

France has interest in three different technologies; the gas cooled fast reactor, the sodium cooled

fast reactor and the very high temperature reactor. France main interest is on the fast reactor is

based on the fact that they will produce less waste and better use the uranium France has. The

governments atomic energy decided to carry on with a generation IV sodium cooled fast breeder

reactor prototype and aimed to start it by 2020. The prototype might be built near the Phenix at

Marcoule and will have an output between 250 to 800 MWe, it is estimated that the cost would

be of about 2 billion EUR. The main objective is to improve the safety and affordability of this

type of reactors, and also to demonstrate advanced recycling methods to improve the ultimate

high-level and long-lived waste to be disposed of. In parallel they are also developing a gas

cooled reactor as an alternative. France expects to have this technology ready for industrial

utilization and for export by 2040 (Nuclear Power in France).

Fast breeder reactor technology has a positive outlook and it is seen as the most viable way of

securing long term energy by various countries. The US has not shown much interest in FBRs

but it is evident that their input and participation is required knowing that this is a way for

acquiring energy, preserving resources and reducing contaminants.


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Future Technologies for Fast Breeder Reactors

During the year 2000 the Generation IV International Forum (GIF) was initiated with the

purpose of developing the next generation of reactors. It is made up of 13 countries, led by the

USA, the members are; Argentina, Brazil, Canada, China, France, Japan, Russia, South Korea,

South Africa, Switzerland and the UK. The GIF focuses on six different technologies of which

three are Fast neutron reactors and one can be built as a fast reactor, one is epithermal and only

two operate similar to todays plants, whit slow neutrons. The fast reactor technologies are; Gas-

cooled fast reactors, Lead-cooled fast reactors, sodium-cooled fast reactors and the Molten salt

fast reactor.

Gas-cooled fast reactors

GFRs will operate at high temperatures; they are suitable for power generation as well as

hydrogen generation. This kind of reactor would have a breeding core with fast neutrons but no

blanket, fuel would include depleted uranium and any other fissile material made into ceramic

pins or plates. Used fuel would be reprocessed on site and all the actinides will be repeatedly

recycled minimizing the long-lived radioactive wastes production. No operating antecedent

exists for the GFR, so a prototype cannot be expected before 2025. EURATOM has planned an

80MWt experimental demonstration by 2014. A lower temperature alternative for GFR exists; it

would be cooled by helium in a primary circuit while supercritical CO2

in a secondary system for

generating power. (Generation IV Nuclear Reactors).

Lead-cooled fast reactors

Corresponding with Russias BREST reactor technology the LFR is a flexible fast

neutron reactor capable of using depleted uranium or thorium as fuel. Cooled by liquid metals

like Pb or Pb-Bi at low pressure, it can also burn actinides from LWR fuel. It is predicted that a

wide range of unit sizes can be accomplished ranging from 300-400 MWe modular units to

single plants of 1400MWe. An operating temperature of 800C is expected with advanced

materials, so lead corrosion resistance can be provided, and thus enabling the thermochemical

production of hydrogen but by now an operating temperature of 550C has already been

achieved. US STAR and Japans LSPR are the base for the design of the LFR. A pool-type

reactor being developed by Toshiba and others in Japan based on the Small Secure Transportable

Autonomous Reactor in the USA. It has an integral steam generator inside the sealed unit. The


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unit is supposed to be mounted underground and a thermal efficiency of 44% is anticipated. 20

MWe versions would have a one meter high core with a diameter of 1.2 meters. Coupled with a

Brayton cycle turbine operated by carbon dioxide and four heat exchangers the SSTAR will

produce power. For recycling the whole unit will be returned after a 20-year life without

refueling.

The European Lead-cooled System of 600 MWe in Europe is led by Ansaldo Nucleare from Italy

and financed by EUTAROM. ELSY neared completion on 2008 and a demonstration facility is

in order. MOX fuel is used and molten lead is pumped to eight steam generators. By 2020 a large

prototype unit is expected and the deployment of small transportable units.

Molten salt reactors (now two variants). In an MSR, the uranium fuel is dissolved in the sodium

fluoride salt coolant which circulates through graphite core channels to achieve some moderation

and an epithermal neutron spectrum. The reference plant is up to 1000 MWe. Fission products

are removed continuously and the actinides are fully recycled, while plutonium and other

actinides can be added along with U-238, without the need for fuel fabrication. Coolant

temperature is 700C at very low pressure, with 800C envisaged. A secondary coolant system is

used for electricity generation, and thermochemical hydrogen production is also feasible.

Compared with solid-fuelled reactors, MSR systems have lower fissile inventories, no radiation

damage constraint on fuel burn-up, no spent nuclear fuel, no requirement to fabricate and handle

solid fuel, and a homogeneous isotopic composition of fuel in the reactor. These and other

characteristics may enable MSRs to have unique capabilities and competitive economics for

actinide burning and extending fuel resources.

During the 1960s the USA developed the molten salt fast reactor as the primary back-up option

for the conventional fast breeder reactor, and a small prototype was operated for about four years.

Recent work has focused on lithium and beryllium fluoride coolant with dissolved thorium and

U-233 fuel. The attractive features of the MSR fuel cycle include: the high-level waste

comprising fission products only, hence shorter-lived radioactivity; small inventory of weapons-


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fissile material (Pu-242 being the dominant Pu isotope); low fuel use (the French self-breeding

variant claims 50kg of thorium and 50kg U-238 per billion kWh); and safety due to passive

cooling up to any size.

For the MSR, no System Arrangements (SA) has been signed, and collaborative R&D is pursued

by interested members under the auspices of a provisional steering committee. There will be a

long lead time to prototypes, and the R&D orientation has changed since the project was set up,

due to increased interest. It now has two baseline concepts:

- The Molten Salt Fast Neutron Reactor (MSFR)

- the Advanced High-Temperature Reactor (AHTR) with the same graphite core structures as the

VHTR and molten salt as coolant instead of helium, enabling power densities 4 to 6 times greater

than HTRs and power levels up to 4000 MWt with passive safety systems.

Sodium-cooled fast reactors. The SFR uses liquid sodium as the reactor coolant, allowing high

power density with low coolant volume. It builds on more than 300 reactor-years experienced

with sodium-cooled fast neutron reactors over five decades and in eight countries, and is the

main technology of interest in GIF. Most plants so far have had a core plus blanket configuration,

but new designs are likely to have all the neutron action in the core. Other R&D is focused on

safety in loss of coolant scenarios, and improved fuel handling.

The SFR utilizes depleted uranium as the fuel matrix and has a coolant temperature of 500-

550C enabling electricity generation via a secondary sodium circuit, the primary one being at

near atmospheric pressure. Three variants are proposed: a 50-150 MWe type with actinides

incorporated into a U-Pu metal fuel requiring electrometallurgical processing (pyro processing)

integrated on site, a 300-1500 MWe pool-type version of this, and a 600-1500 MWe type with

conventional MOX fuel and advanced aqueous reprocessing in central facilities elsewhere.

Early in 2008, the USA, France and Japan signed an agreement to expand their cooperation on

the development of sodium-cooled fast reactor technology. The agreement relates to their

collaboration in the Global Nuclear Energy Partnership, aimed at closing the nuclear fuel cycle


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through the use of advanced reprocessing and fast reactor technologies, and seeks to avoid

duplication of effort.

future of fast breeder reactors final

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