A

Seminar Report Entitled on

““AADDVVAANNCCEE NNUUCCLLEEAARR RREEAACCTTOORRSS””

INDEX

SR.NO CHAPTER PAGE NO. 1 Introduction 1

2 Current technologies in nuclear reactor 2.1 Pressurised Water Reactor 2

2.2 Boiling Water Reactor 3 2.3 Pressurised Heavy Water Reactor 4

2.4 Gas Cooled Reactor 4 2.5 Liquid Metal Fast Breeder Reactor 5

2.5.1 Lead Cooled 6 2.5.2 Sodium Cooled 7 2.6 Pebble Bed Reactor 7

2.7 Molten Salt Reactor 8

3 Future and advancing techniques of reactor 3.1 Integrated Fast Reactor 9

3.2 Pebble Bed Reactor 9 3.3 Small, Sealed, Transportable Autonomous

Reactor 10 3.4 Clean And Environmental Safe Advanced

Reactor 10 3.5 Reduced Moderated Water Reactor 10

3.6 Hydrogen Moderated Self Regulating Nuclear Power Module 10

3.7 Subcritical Reactor 11 3.8 Thorium Based Reactors 11 3.8.1 Advanced Heavy Water Reactor 11

4 Generation IV Reactor

4.1 Gas Cooled Fast Reactor 12 4.2 Lead Cooled Fast Reactor 13

4.2.1 Advantages 14 4.2.2 Disadvantages 14

4.3 Molten Salt Reactor 15 4.3.1 Advantages 16

4.3.2 Disadvantages 16

5 Conclusion 17

References 18

List of Figures

Fig no. Name of the figure Page no.

2.1 Pressurised Water Reactor PWR 2

2.2 Pressurised Water Vessel Head 3

2.3 Boiling Water Reactor BWR 3

2.4 Pressurised Heavy Water Reactor 4

2.5 Gas Cooled Reactor 5

2.6 Liquid Metal Fast Breeder Reactor 6

2.7 Pebble Bed Reactor 7

2.8 Molten Salt Reactor 8

3.1 Advanced Pebble Bed Reactor 9

3.2 Hydrogen Moderated Power Module Reactor 11

4.1 Gas Cooled Fast Reactor 12

4.2 Lead Cooled Fast Reactor 13

4.3 Molten Salt Reactor 15

CHAPTER 1

INTRODUCTION

A nuclear reactor is a device to initiate and control a sustained nuclear chain reaction.

Nuclear reactors are used at nuclear power plants for electricity generation and in propulsion of

ships. Heat from nuclear fission is passed to a working fluid (water or gas), which runs

through turbines. These either drive a ship's propellers or turn electrical generators. Nuclear

generated steam in principle can be used for industrial process heat or for district heating. Some

reactors are used to produce isotopes for medical and industrial use, or for production

of plutonium for weapons. Some are run only for research. Today there are about 450 nuclear

power reactors that are used to generate electricity in about 30 countries around the world.

The nuclear power industry has been developing and improving reactor technology for

more than five decades and is starting to build the next generation of nuclear power reactors to

fill new orders.

Several generations of reactors are commonly distinguished. Generation I reactors were

developed in 1950-60s, and outside the UK none are still running today. Generation II reactors

are typified by the present US and French fleets and most in operation elsewhere. So-called

Generation III (and 3+) are the Advanced Reactors discussed in this paper, though the distinction

from Generation II is arbitrary. The first are in operation in Japan and others are under

construction or ready to be ordered. Generation IV designs are still on the drawing board and

will not be operational before 2020 at the earliest.

About 85% of the world's nuclear electricity is generated by reactors derived from designs

originally developed for naval use. These and other nuclear power units now operating have been

found to be safe and reliable, but they are being superseded by better designs.

Reactor suppliers in North America, Japan, Europe, Russia and elsewhere have a dozen

new nuclear reactor designs at advanced stages of planning.

CHAPTER 2

CURRENT TECHNOLOGIES IN NUCLEAR REACTOR

2.1 PRESSURISED WATER REACTOR (PWR)

fig.2.1 pressurised water reactor PWR

These reactors use a pressure vessel to contain the nuclear fuel, control rods, moderator,

and coolant. They are cooled and moderated by high-pressure liquid water. The hot radioactive

water that leaves the pressure vessel is looped through a steam generator, which in turn heats a

secondary (non-radioactive) loop of water to steam that can run turbines. They are the majority

of current reactors. This is a thermal neutron reactor design, the newest of which are the

VVER-1200, Advanced Pressurized Water Reactor and the European Pressurized

Reactor. United States Naval reactors are of this type.

Pressurized water reactors (PWRs) constitute the large majority of all Western nuclear

power plants and are one of three types of light water reactor (LWR), the other types being

boiling water reactors (BWRs) and supercritical water reactors (SCWRs). In a PWR, the primary

coolant (water) is pumped under high pressure to the reactor core where it is heated by the

energy generated by the fission of atoms. The heated water then flows to a steam generator

where it transfers its thermal energy to a secondary system where steam is generated and flows to

turbines which, in turn, spin an electric generator.

PWRs were originally designed to serve as nuclear propulsion for nuclear submarines and

were used in the original design of the second commercial power plant at Shipping port Atomic

Power Station.

fig.2.2 pressurised water vessel head

2.2 BOILING WATER REACTOR (BWR)

fig.2.3 boiling water reactor BWR

A BWR is like a PWR without the steam generator. A boiling water reactor is cooled and

moderated by water like a PWR, but at a lower pressure, which allows the water to boil inside

the pressure vessel producing the steam that runs the turbines. Unlike a PWR, there is no primary

and secondary loop. The thermal efficiency of these reactors can be higher, and they can be

simpler, and even potentially more stable and safe. This is a thermal neutron reactor design, the

newest of which are the Advanced Boiling Water Reactor and the Economic Simplified

Boiling Water Reactor.

2.3 PRESSURISED HEAVY WATER REACTOR

fig.2.4 Pressurised Heavy Water Reactor

A Canadian design (known as CANDU), these reactors are heavy-water-cooled and -

moderated pressurized-water reactors. Instead of using a single large pressure vessel as in a

PWR, the fuel is contained in hundreds of pressure tubes. These reactors are fuelled with natural

uranium and are thermal neutron reactor designs. PHWRs can be refuelled while at full power,

which makes them very efficient in their use of uranium (it allows for precise flux control in the

core). CANDU PHWRs have been built in Canada, Argentina, China, India, Pakistan,

Romania, and South Korea. India also operates a number of PHWRs, often termed 'CANDU-

derivatives', built after the Government of Canada halted nuclear dealings with India following

the 1974 Smiling Buddha nuclear weapon test.

2.4 GAS-COOLED REACTOR (GCR) AND ADVANCED GAS-COOLED

REACTOR (AGR)

These are generally graphite moderated and CO2 cooled. They can have a high thermal

efficiency compared with PWRs due to higher operating temperatures. There are a number of

operating reactors of this design, mostly in the United Kingdom, where the concept was

developed.

fig.2.5 Gas Cooled Reactor

Older designs (i.e. Mango stations) are either shut down or will be in the near future. However,

the AGCRs have an anticipated life of a further 10 to 20 years. This is a thermal neutron reactor

design. Decommissioning costs can be high due to large volume of reactor core.

There were two main types of generation I GCR

1. The Magnox reactors developed by the United Kingdom

2.The UNGG reactors developed by France.

The main difference between these two types is in the fuel cladding material. Both types

were mainly constructed in their countries of origin, with a few export sales: Magnox plants to

Italy and Japan, and a UNGG to Spain. Both types used fuel cladding materials that were

unsuitable for medium term storage under water, making reprocessing an essential part of the

cycle. Both types were also designed and used to produce weapons-grade plutonium, but at the

cost of major interruption to their use for power generation despite the provision of online

refuelling.

2.5 LIQUID METAL FAST BREEDER REACTOR (LMFBR)

This is a reactor design that is cooled by liquid metal, totally unmoderated, and produces

more fuel than it consumes.

fig.2.6 Liquid Metal Fast Breeder Reactor

They are said to "breed" fuel, because they produce fissionable fuel during operation

because of neutron capture. These reactors can function much like a PWR in terms of

efficiency, and do not require much high-pressure containment, as the liquid metal does not need

to be kept at high pressure, even at very high temperatures. BN-350 and BN-600 in USSR and

Superphénix in France were a reactor of this type, as was Fermi-I in the United States. The

Monju reactor in Japan suffered a sodium leak in 1995 and was restarted in May 2010. All of

them use/used liquid sodium. These reactors are fast neutron, not thermal neutron designs.

These reactors come in two types:

2.5.1 LEAD COOLED

Using lead as the liquid metal provides excellent radiation shielding, and allows for

operation at very high temperatures. Also, lead is (mostly) transparent to neutrons, so fewer

neutrons are lost in the coolant, and the coolant does not become radioactive. Unlike sodium,

lead is mostly inert, so there is less risk of explosion or accident, but such large quantities of lead

may be problematic from toxicology and disposal points of view. Often a reactor of this type

would use a lead-bismuth eutectic mixture. In this case, the bismuth would present some

minor radiation problems, as it is not quite as transparent to neutrons, and can be transmuted to a

radioactive isotope more lead. The Russian Alfa class submarine uses a lead-bismuth-cooled

fast reactor.

2.5.2 SODIUM COOLED

Most LMFBRs are of this type. The sodium is relatively easy to obtain and work with,

and it also manages to actually prevent corrosion on the various reactor parts immersed in it.

However, sodium explodes violently when exposed to water, so care must be taken, but such

explosions would not be vastly more violent than (for example) a leak of superheated fluid from

a SCWR or PWR. EBR-I, the first reactor to have a core meltdown, was of this type.

2.6 PEBBLE-BED REACTORS

fig.2.7 Pebble Bed Reactor

These use fuel moulded into ceramic balls, and then circulate gas through the balls. The result is

an efficient, low-maintenance, very safe reactor with inexpensive, standardized fuel. The

prototype was the AVR.

2.7 MOLTEN SALT REACTORS

These dissolve the fuels in fluoride salts, or use fluoride salts for coolant. These have

many safety features, high efficiency and a high power density suitable for vehicles.

fig.2.8 Molten Salt Reactor

Notably, they have no high pressures or flammable components in the core. The prototype was

the MSRE, which also used Thorium's fuel cycle to produce 0.1% of the radioactive waste of

standard reactor.

CHAPTER 3

FUTURE AND DEVELOPING TECHNOLOGIES

ADVANCED REACTOR

3.1 INTEGRATED FAST REACTOR (IFR)

The Integral Fast Reactor (IFR) was built, tested and evaluated during the 1980s and then

retired under the Clinton administration in the 1990s due to nuclear non-proliferation policies of

the administration. Recycling spent fuel is the core of its design and it therefore produces only a

fraction of the waste of current reactors.

3.2 PEBBLE BED REACTOR

fig.3.1 Advanced Pebble Bed Reactor

The pebble-bed reactor, a high-temperature gas-cooled reactor (HTGCR), is designed so

high temperatures reduce power output by Doppler broadening of the fuel's neutron cross-

section.

It uses ceramic fuels so its safe operating temperatures exceed the power-reduction

temperature range. Most designs are cooled by inert helium. Helium is not subject to steam

explosions, resists neutron absorption leading to radioactivity, and does not dissolve

contaminants that can become radioactive.

Typical designs have more layers (up to 7) of passive containment than light water

reactors (usually 3). A unique feature that may aid safety is that the fuel-balls actually form the

core's mechanism, and are replaced one-by-one as they are. The design of the fuel makes fuel

reprocessing expensive.

3.3 SMALL, SEALED, TRANSPORTABLE AUTONOMOUS REACTOR

The Small, sealed, transportable, autonomous reactor (SSTAR) is being primarily

researched and developed in the US, intended as a fast breeder reactor that is passively safe and

could be remotely shut down in case the suspicion arises.

3.4 CLEAN AND ENVIRONMENTAL SAFE ADVANCED REACTOR (CAESAR)

The Clean And Environmentally Safe Advanced Reactor (CAESAR) is a nuclear reactor

concept that uses steam as a moderator – this design is still in development.

3.5 REDUCED MODERATION WATER REACTOR

The Reduced moderation water reactor builds upon the Advanced boiling water

reactor(ABWR) that is presently in use, it is not a complete fast reactor instead using mostly

epithermal neutrons, which are between thermal and fast neutrons in speed

3.6 HYDROGEN MODERATED SELF REGULATING NUCLEAR POWER

MODULE(HPM)

The hydrogen-moderated self-regulating nuclear power module (HPM) is a reactor design

emanating from the Los Alamos National Laboratory that uses uranium hydride as fuel.

fig.3.2 Hydrogen Moderated Power Module Reactor

3.7 SUBCRITICAL REACTORS

These are designed to be safer and more stable, but pose a number of engineering and

economic difficulties. One example is the Energy amplifier

3.8 THORIUM BASED REACTORS

Thorium-based reactors. It is possible to convert Thorium-232 into U-233 in reactors

specially designed for the purpose. In this way, thorium, which is more plentiful than uranium,

can be used to breed U-233 nuclear fuel. U-233 is also believed to have favourable nuclear

properties as compared to traditionally used U-235, including better neutron economy and lower

production of long lived transuranic waste.

3.8.1 ADVANCED HEAVY WATER REACTOR (AHWR)

(AHWR)— A proposed heavy water moderated nuclear power reactor that will be the next

generation design of the PHWR type. Under development in the BHABHA Atomic Research

Centre (BARC), India.

CHAPTER 4

GENERATION IV REACTORS

Generation IV are a set of theoretical nuclear reactor designs currently being researched.

These designs are generally not expected to be available for commercial construction before

2030. Current reactors in operation around the world are generally considered second- or third-

generation systems, with the first-generation systems having been retired some time ago.

Research into these reactor types was officially started by the Generation IV International Forum

(GIF) based on eight technology goals. The primary goals being to improve nuclear safety,

improve proliferation resistance, minimize waste and natural resource utilization, and to decrease

the cost to build and run such plants.

TYPES OF GENERATION IV REACTOR

4.1 Gas Cooled Fast Reactor

4.2Lead Cooled Fast Reactor

4.3Molten Salt Reactor

4.4.Sodium Cooled Fast Reactor

4.5Very High Temperature Reactor

4.1 GAS COOLED FAST REACTOR

fig.4.1 Gas Cooled Fast Reactor

The gas-cooled fast reactor (GFR) system is a nuclear reactor design which is currently in

development. Classed as a Generation IV reactor, it features a fast-neutron spectrum and closed

fuel cycle for efficient conversion of fertile uranium and management of actinides. The reference

reactor design is a helium-cooled system operating with an outlet temperature of 850 °C using a

direct Brayton closed-cycle gas turbine for high thermal efficiency. Several fuel forms are being

considered for their potential to operate at very high temperatures and to ensure an excellent

retention of fission products composite ceramic fuel, advanced fuel particles, or ceramic clad

elements of actinide compounds. Core configurations are being considered based on pin- or

plate-based fuel assemblies or prismatic blocks, which allows for better coolant circulation than

traditional fuel assemblies.

The reactors are intended for use in nuclear power plants to produce electricity, while at

the same time producing (breeding) new nuclear fuel.

4.2 LEAD COOLED FAST REACTOR

fig.4.2 Lead Cooled Fast Reactor

Molten lead or lead-bismuth eutectic can be used as the primary coolant in a nuclear

reactor, because lead and bismuth have low neutron absorption and relatively low melting points.

Neutrons are slowed less by interaction with these heavy nuclei, (thus not being neutron

moderators) and therefore help make this type of reactor a fast-neutron reactor. The coolant does

however serve as a neutron reflector, returning some escaping neutrons to the core.

Few have been constructed, including some Soviet nuclear submarine reactors in the

1970s, but a number of proposed new nuclear reactor designs are lead cooled. Some designs

are claimed to be able to circulate the primary coolant via convection without requiring pumps,

at least in emergency shutdown conditions.

The Gen IV lead-cooled fast reactor is a nuclear reactor that features a fast neutron

spectrum, molten lead or lead-bismuth eutectic coolant. Options include a range of plant ratings,

including a number of 50 to 150 MWe (megawatts electric) units featuring long- life, pre-

manufactured cores. Plans include modular arrangements rated at 300 to 400 MWe, and a large

monolithic plant rated at 1,200 MWe. The fuel is metal or nitride-based containing fertile

uranium and transuranics. A smaller capacity LFR such as SSTAR can be cooled by natural

convection, larger proposals such as ELSY use forced circulation in normal power operation, but

with natural circulation emergency cooling. The reactor outlet coolant temperature is typically in

the range of 500 to 600 °C, possibly ranging over 800 °C with advanced materials for later

designs. Temperatures higher than 800 °C are high enough to support thermochemical

production of hydrogen.

4.2.1 ADVANTAGES

Instead of refuelling, the whole core can be replaced after many years of operation. Such a

reactor is suitable for countries that do not plan to build their own nuclear infrastructure.

Lifetime without refuelling can be increased more easily, in part due to higher efficiency

As no electricity is required for the cooling after shutdown, this design has the potential to be

safer than a water-cooled reactor.

4.2.2 DISADVANTAGES

Solidification of the lead-bismuth solution renders the reactor inoperable. However, lead-

bismuth eutectic has a comparatively low melting temperature of 123.5°C (254.3°F), making

desolidification a relatively easily accomplished task. Lead has a higher melting point of

327.5°С, but is often used as a pool type reactor where the large bulk of lead does not easily

freeze.

4.3 MOLTEN SALT REACTOR

fig.4.3 Molten Salt Reactor

A molten salt reactor (MSR) is a class of nuclear fission reactors in which the primary

coolant, or even the fuel itself, is a molten salt mixture. MSRs run at higher temperatures than

water-cooled reactors for higher thermodynamic efficiency, while staying at low vapour

pressure.

The nuclear fuel may be solid or dissolved in the coolant itself. In many designs the

nuclear fuel is dissolved in the molten fluoride salt coolant a uranium tetrafluoride (UF4). The

fluid becomes critical in a graphite core which serves as the moderator. Solid fuel designs rely on

ceramic fuel dispersed in a graphite matrix, with the molten salt providing low pressure, high

temperature cooling. The salts are much more efficient than compressed helium at removing heat

from the core, reducing the need for pumping and piping and reducing the size of the core.

The early Aircraft Reactor Experiment (1954) was primarily motivated by the small size

that the design could provide, while the Molten-Salt Reactor Experiment (1965–1969) was a

prototype for a thorium fuel cycle breeder reactor nuclear power plant. One of the Generation IV

reactor designs is a molten-salt-cooled, molten-salt- fuelled reactor (illustrated on the right); the

initial reference design is 1000MWe.

4.3.1 ADVANTAGES

Inherently safe design (safety by passive components and the strong negative temperature

coefficient of reactivity of some designs).

Operating at a low pressure improves safety and simplifies the design

In theory a full recycle system can be much cleaner: the discharge wastes after chemical

separation are predominately fission products, most of which have relatively short half lives

compared to longer- lived actinide wastes. This can result in a significant reduction in the

containment period in a geologic.

4.3.2 DISADVANTAGES

Little development compared to most Gen IV designs much is unknown.

CHAPTER 5

CONCLUSION

By using advanced nuclear reactors, the efficiency of nuclear power plant can be

increased. The primary goals being to improve nuclear safety, improve proliferation resistance,

minimize waste and natural resource utilization and to decrease the cost to build and run such

plants.

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