Recent Trends in Nuclear Power Plant

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CHAPTER 1

INTRODUCTION TO NUCLEAR ENERGY

1.1 Background An atom has a large amount of energy holding their nuclei

together. Certain isotopes of some elements can be split and will release energy. Part of

their energy as heat. This splitting is called fission. The heat released in fission can be

used to generate electricity in nuclear power plants.

Uranium-235 (U-235) is one of the isotopes that fissions easily. During fission, U-

235 atoms absorb loose neutrons. This causes U-235 to become unstable and split into

two light atoms called fission products. The combined mass of the fission products is lessthan that of the original U-235. The reduction occurs because some of the matter changes

into energy. The energy is released as heat. Two or three neutrons are released along with

the heat. These neutrons may hit other atoms, causing more fission. Thus energy released

in form of heat is used to generate power in nuclear power plant.

The key to capacity nuclear power sector critically depends upon development of

reliable nuclear power station, encompassing complete fuel breeding. So far India has

developed to give three stage nuclear program of Dr. Homi J. Bhabha. The first stageutilizes natural uranium as fuel and heavy water as moderator in pressurized heavy water

reactor. Second stage are based on fast breeder reactors (FBR) which are fuelled by

plutonium obtained by reprocessing of spent fuel of thermal reactors. Fast reactor

produces more fissile material than what they consume and thus multiplication of fissile

inventory and enhancement of installed capacity. Third stage will based on thorium U 233

cycle. Timely implementation of this stage is very crucial for meeting increasing carbon

free energy demand in country. As we have large resource of thorium it is our duty to

design reactor which will implement effectively thorium fuel cycle. Thus research at

BARC developed design of advanced heavy water reactor (AHWR).We have to focus on

auxiliary application, efficiency and innovative technology of nuclear power [1].

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1.2History Ancient Greek philosophers first developed the idea that all

matter is composed of invisible particles called atom. By 1900, physicists knew the

atom contains large quantities of energy. British physicist Ernest Rutherford was

called the father of nuclear science because of his contribution to the theory of atomic

structure. In 1904 he wrote, if it were ever possible to control rate of disintegration of

the radio elements, an enormous amount of energy could be obtained from a small

amount of matter. Albert Einstein developed his theory of the relationship between

mass and energy one year later. The mathematical relation is

(1.1)

E=energy released J/s

m=mass of atom kg

C=velocity of light =3108 m/s

Few milestones in history of development of nuclear power generation are as below.

1.2.1 In the decade 40s December 2, 1942. The first self-sustaining nuclear

chain reaction occurs at the University of Chicago.

July 16, 1943. The U.S. Armys Manhattan Engineer District (MED) tests the first

atomic bomb at Alamogordo, New Mexico, under the code me Manhattan Project.

August 6, 1945. The atomic bomb nicknamed little boy is dropped on Hiroshima,

Japan. Three days later, another bomb, Fat man, is dropped on Nagasaki, Japan. Japan

surrenders on August 15, ending World War II.

August 1, 1946. The Atomic Energy Act of 1946 creates the Atomic Energy

Commission (AEC) to control nuclear energy development and explore peaceful uses

of nuclear energy.

October 6, 1947. The AEC first investigates the possibility of peaceful uses of

atomic energy, issuing a report the following year.

1.2.2 In the decade 50s December 20, 1951. In Arco, Idaho, Experimental

Breeder Reactor I produces the first electric power from nuclear energy, lighting four

light bulbs.

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June 14, 1952. Keel for the Navy's first nuclear submarine, Nautilus, is laid at

Groton, Connecticut.

March 30, 1953. Nautilus starts its nuclear power units for the first time.

August 30, 1954. President Eisenhower signs The Atomic Energy Act of 1954, the

first major meant of the original Atomic Energy Act, giving the civilian nuclear

power program further access to nuclear technology.

January 10, 1955. The AEC announces the power demonstration Reactor Program.

Under the program, AEC and industry will cooperate in constructing and operating

experimental nuclear power reactors.

1.2.3 In the decade 60s August 19, 1960. The third U.S. nuclear power

plant, Yankee Rowe Nuclear Power Station, achieves a self-sustaining nuclear

reactor.

Early 1960s. Small nuclear-power generators are first used in remote areas to power

weather stations and to light buoys for sea navigation.

1.2.4 In the decade 70s 1970 March 5. The United States, United

Kingdom, Soviet Union, and 45 other nations ratify the Treaty for Non-Proliferation

of Nuclear Weapons.

1971 Twenty-two commercial nuclear power plants are in full operation in the United

States. They produce 2.4 percent of U.S. electricity at this time.

1973 U.S. utilities order 41 nuclear power plants, a one-year record.

1974 The first 1,000-megawatt-electric nuclear power plant goes into service

Commonwealth Edisons Zion 1 plant.

August 4, 1977. President Carter signs the Department of Energy Organization Act,

which transfers ERDA functions to the new.

The nuclear power industry in the U.S. grew rapidly in the 1960s. In the

1970s and 1980s, however, growth slowed. Demand for electricity decreased and

concern grew over nuclear issues, such as reactor safety, waste disposal, and other

environmental consideration. At the end of 1991, 31 other countries also had nuclear

power plants in commercial operation or under construction.

1.2.5 Milestones in development of nuclear power in India

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March 12, 1944 Dr. Homi Jehangir Bhabha writes to Sir Dorabji Tata Trust for

starting Nuclear Research in India

December 19, 1945 Tata Institute of Fundamental Research (TIFR) Mumbai is

inaugurated.

April 15, 1948 Atomic Energy Act is passed

August 04, 1956 Apsara first research reactor in Asia attains criticality at Tomboy,

Mumbai.

January 20, 1957 Atomic Energy Establishment, Trombay (AEET) is started

May 10, 1984 Research Reactor PURNIMA-II, a Uranium-233 fuelled homogenous

reactor, attains criticality.

September 17, 1987 Nuclear Power Corporation of India Limited (NPCIL) is formed

by converting the erstwhile Nuclear Power Board.

May 11 & 13, 1998 Five underground nuclear tests are conducted at Pokhran Range,

Rajasthan.

October 31, 2002 Waste Immobilization Plant and Uranium-Thorium Separation

Plant at (both at Tromboy), and the Radiation Processing Plant Krushak at Lasalgaon,

district Nasik, Maharashtra, are dedicated to the Nation.

August 31, 2007 Units 3 and 4 of the Tarapur Atomic Power Station dedicated to the

Nation.

1.3 Basics of nuclear power plant The purpose of a nuclear power plant

is to produce electricity. It should also be obvious that nuclear power plants have

some significant differences from other plants. In a nuclear power plant, many of the

components are similar to those in a fossil-fueled plant, except that the steam boiler is

replaced by a Nuclear Steam Supply System (NSSS). The NSSS consists of a nuclear

reactor and all of the components necessary to produce high pressure steam, which

will be used to turn the turbine for the electrical generator. Like a fossil-fueled plant,

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a nuclear power plant boils water to produce electricity. Unlike a fossil-fueled plant,

the nuclear plants energy does not come from the combustion of fuel, but from the

fissioning (splitting) of fuel atoms [1].

Fig 1.1 Nuclear power generation.

1.3.1 Fission When high velocity neutron strikes other heavy nucleus at rest it

splits into two or more light nuclei. Unlike a fossil-fueled plant, the nuclear plants

energy does not come from the combustion of fuel, but from the fissioning (splitting) of

fuel atoms. The most common fuel for the electrical producing reactor plants is uranium.

Amount of energy or heat is evolved during fission reaction. Energy release depends on

type of reaction. While reaction give rise to fission fragments radioactive isotope, fissile

atom fast moving neutron for further reaction.

92U235 + 0n

1 56Ba137 +36Kr97 +20n

1 (1.2)

235.0439 + 1.0087 136.906 + 96.921+ 2*1.00867 (1.3)

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236.0526 235.8446 (1.4)

... m= - 0.2080 amu. And

... E =931*- 0.2080 = - 193 Me

Fig 1.2 chain reaction of fission of Uranium -235.

Fig 1.2 tells us that when U235 is collided by free slow moving neutron it gets

splited into fission fragment (Ba. and Kr.) with reduced mass as that of original nucleus.While doing so it reduces energy which is indirectly used for power generation. Chain

reaction equation above tells us that quantity of energy can be generated for give set of

fuel and chain reaction.

1.3.2 Fusion Four nuclei of hydrogen combine together to fuse in series .It gives

combined fused product of two helium atom. During fusion very high temperature.

Hundred million degree has to archive thus large energy is released as compared tofission .Plasma is formed e .g .Sun nucleus exothermic reaction, gain in mass.

41H1

2He2 + 2 1e

0 (1.5)

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1.3.3 Enrichment The most common fuel for the electrical producing

reactor plants in the United States is uranium. The uranium starts out as ore, and

contains a very low percentage (or low enrichment) of the desired atoms (U-235). The

U-235 is a more desirable atom for fuel, because it is easier to cause the U-235 atoms to

fission (split) than the much more abundant U-238 atoms. Therefore, the fuel

fabrication process includes steps to increase the number of U-235 atoms in relation to

the number of U-238 atoms (enrichment process).

i. ii.

Fig 1.3 fuel uranium i) raw ii) pallet

1.3.4 Reactor Fuel Assembly 1.3.4.1 Both boiling water reactor

and pressurized water reactor fuel assemblies consist of the same major components.

These major components are the fuel rods, the spacer grids, and the upper and lower

end fittings. The fuel assembly drawing fig.1.4 shows these major components

(pressurized water reactor fuel assembly).

1.3.4.2 The fuel rods contain the ceramic fuel pellets. The fuel rods are

approximately 12 feet long and contain a space at the top for the collection of any

gases that are produced by the fission process. These rods are arranged in a square

matrix ranging from 17 x 17 for pressurized water reactors to 8 x 8 for boiling water

reactors.

1.3.4.3 The spacer grids separate the individual rods with pieces of sprung metal.

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This provides the rigidity of the assemblies and allows the coolant to flow freely up

through the assemblies and around the fuel rods. Some spacer grids may have flow

mixing vanes that are used to promote mixing of the coolant as it flows around and

though the fuel assembly.

1.3.4.4 The upper and lower end fittings serve as the upper and lower structural

elements of the assemblies. The lower fitting (or bottom nozzle) will direct the

coolant flow to the assembly through several small holes machined into the fitting.

There are also holes drilled in the upper fitting to allow the coolant flow to exit the

fuel assembly.

1.3.4.5 The upper end fitting will also have a connecting point for the refueling

equipment to attach for the moving of the fuel with a crane. For pressurized water

reactor fuel, there will also be guide tubes in which the control rods travel operations.

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Fig. 1.4 Reactor core assembly.

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1.4 Basic types of reactor

1.4.1 Boiling water reactor There are two basic types of reactor plants

being used in the United States to produce electricity, the boiling water reactor

(BWR) and the pressurized water reactor (PWR). The boiling water reactor operates

in essentially the same way as a fossil-fueled generating plant

Fig. 1.5 Boiling water reactor power generation.

Inside the reactor vessel, a steam/water mixture is produced when very

pure water (reactor coolant) moves upward through the ore absorbing heat. The

major difference in the operation of a boiling water reactor as compared to other

nuclear systems is the steam void formation in the core. The steam/water mixture

leaves the top of the core and enters two stages of moisture separation, where

water droplets are removed before the steam is allowed to enter the steam line.

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The steam line, in turn, directs the steam to the main turbine, causing it to turn the

turbine and the attached electrical generator. The unused steam is exhausted to

the condenser where it is condensed into water. The resulting water (condensate)

is pumped out of the condenser with a series of pumps and back to the reactor

vessel. The recirculation pumps and the jet pumps allow the operator to vary

coolant flow through the core and to change reactor power.

1.4.2 Pressurized water reactor The pressurized water reactor (PWR)

differs from the boiling water reactor in that steam is produced in the steam

generator rather than in the reactor vessel. The pressurizer keeps the water that is

flowing through the reactor vessel under very high pressure (> 2,200 pounds per

square inch) to prevent it from boiling, even at operating temperatures of more than

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600EF.

Fig. 1.6 Pressurized water reactor

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, spins an

electric generator. In contrast to a boiling water reactor, pressure in the primary coolant

loop prevents the water from boiling within the reactor. All LWRs use ordinary light

water as both coolant and neutron and moderator. PWRs were originally designed to

serve as nuclear submarinepower plants and were used in the original design of the

second commercial power plant atShipping port Atomic Power Station. PWRs currently

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operating in the United States are considered Generation II reactors. Russia's VVER

reactors are similar to U.S. PWRs. France operates many PWRs to generate the bulk of

their electricity.

1.4.2.1 PWR reactor design Nuclear fuel in the reactor vessel is engaged

in afission chain reaction, which produces heat, heating the water in the primary coolant

loop by thermal conduction through the fuel cladding. The hot primary coolant is pumped

into a heat exchanger called the steam generator, where it flows through hundreds or

thousands of tubes (usually 3/4 inch in diameter). Heat is transferred through the walls of

these tubes to the lower pressure secondary coolant located on the sheet side of the

exchanger where it evaporates to pressurized steam. The transfer of heat is accomplished

without mixing the two fluids, which is desirable since the primary coolant might becomeradioactive. Some common steam generator arrangements are u-tubes or single pass heat

exchangers

In a nuclear power station, the pressurized steam is fed through a steam turbine

which drives an electrical generatorconnected to the electric grid for distribution. After

passing through the turbine the secondary coolant (water-steam mixture) is cooled down

and condensed in a condenser. The condenser converts the steam to a liquid so that it can

be pumped back into the steam generator, and maintains a vacuum at the turbine outlet so

that the pressure drop across the turbine, and hence the energy extracted from the steam,

is maximized. Before being fed into the steam generator, the condensed steam (referred

to as feed water) is sometimes preheated in order to minimize thermal shock.

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Fig 1.7. Design of pressurized water reactor

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The steam generated has other uses besides power generation. In nuclear ships

and submarines, the steam is fed through a steam turbine connected to a set of speed

reduction gearsto a shaft used forpropulsion. Direct mechanical action by expansion of

the steam can be used for a steam-powered aircraft catapult or similar applications.

District heating by the steam is used in some countries and direct heating is applied to

internal plant applications. Two things are characteristic for the pressurized water reactor

(PWR) when compared with other reactor types: coolant loop separation from the steam

system and pressure inside the primary coolant loop.

In a PWR, there are two separate coolant loops (primary and secondary), which

are both filled with mineralized/deionizer water. A boiling water reactor, by contrast, has

only one coolant loop, while more exotic designs such as breeder reactors use substancesother than water for coolant and moderator (e.g. sodium in its liquid state as coolant or

graphite as a moderator). The pressure in the primary coolant loop is typically 1516

Mpa (150160 bar), which is notably higher than in other nuclear reactors, and nearly

twice that of a boiling water reactor (BWR). As an effect of this, only localized boiling

occurs and steam will re-condense promptly in the bulk fluid. By contrast, in a boiling

water reactor the primary coolant is designed to boil PWR reactor vessel

1.4.2.2 Coolant Light wateris used as the primary coolant in a PWR. It enters the

bottom of the reactor core at about 275 C (530F) and is heated as it flows upwards

through the reactor core to a temperature of about 315 C (600 F). The water remains

liquid despite the high temperature due to the high pressure in the primary coolant loop,

usually around 155bar(15.5 MPa 153atm, 2,250psig). In water, the critical point occurs

at around 647 K(374 C or 705F) and 22.064 MPa (3200 PSIA or 218 atm)

Pressure in the primary circuit is maintained by a pressurizer, a separate vessel

that is connected to the primary circuit and partially filled with water which is heated to

the saturation temperature (boiling point) for the desired pressure by submerged electrical

heaters. To achieve a pressure of 155 bar, the pressurizer temperature is maintained at

345 C, which gives a sub cooling margin (the difference between the pressurizer

temperature and the highest temperature in the reactor core) of 30 C. Thermal transients

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in the reactor coolant system result in large swings in pressurizer liquid volume, total

pressurizer volume is designed around absorbing these transients without uncovering the

heaters or emptying the pressurizer. Pressure transients in the primary coolant system

manifest as temperature transients in the pressurize and are controlled through the use of

automatic heaters and water spray, which raise and lower pressurize temperature,

respectively. To achieve maximum heat transfer, the primary circuit temperature,

pressure and flow rate are arranged such that sub cooled nucleate boilingtakes place as

the coolant passes over the nuclear fuel rods.

The coolant is pumped around the primary circuit by powerful pumps, which can

consume up to 6 MW each[5]. After picking up heat as it passes through the reactor core,

the primary coolant transfers heat in a steam generator to water in a lower pressuresecondary circuit, evaporating the secondary coolant to saturated steam in most designs

6.2 MPa (60 atm, 900psia), 275 C (530 F) for use in the steam turbine. The cooled

primary coolant is then returned to the reactor vessel to be heated again.

1.4.2.3 Moderator Pressurized water reactors, like thermal reactor designs,

require the fast fission neutrons to be slowed down (a process called moderation or

thermalization) in order to interact with the nuclear fuel and sustain the chain reaction. In

PWRs the coolant water is used as a moderator by let, the neutrons undergo multiple

collisions with light hydrogen atoms in the water, losing speed in the process. This

"moderating" of neutrons will happen more often when the water is denser (more

collisions will occur). The use of water as a moderator is an important safety feature of

PWRs, as an increase in temperature may cause the water to turn to steam - thereby

reducing the extent to which neutrons are slowed down and hence reducing the reactivity

in the reactor. Therefore, if reactivity increases beyond normal, the reduced moderation

of neutrons will cause the chain reaction to slow down, producing less heat. This property, known as the negative temperature coefficient of reactivity, makes PWR

reactors very stable.

In contrast, the RBMK reactor design used at Chernobyl, which uses graphite

instead of water as the moderator and uses boiling water as the coolant, has a large

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positive thermal coefficient of reactivity, which increases heat generation when coolant

water temperatures increase. This makes the RBMK design less stable than pressurized

water reactors. In addition to its property of slowing down neutrons when serving as a

moderator, water also has a property of absorbing neutrons, albeit to a lesser degree.

When the coolant water temperature increases, the boiling increases, which creates voids.

Thus there is less water to absorb thermal neutrons that have already been slowed down

by the graphite moderator, causing an increase in reactivity. This property is called the

void coefficient of reactivity, and in an RBMK reactor like Chernobyl, the void

coefficient is positive, and fairly large, causing rapid transients. This design characteristic

of the RBMK reactor is generally seen as one of several causes of the Chernobyl

accident[6].

Heavy water has very low neutron absorption, so heavy water reactors such as

CANDU reactors also have a positive void coefficient, though it is not as large as that of

an RBMK like Chernobyl; these reactors are designed with a number of safety systems

not found in the original RBMK design, which are designed to handle or react to this as

needed.

PWRs are designed to be maintained in an under moderated state. Meaning that

there is room for increased water volume or density to further increase moderation,

because if moderation were near saturation, then a reduction in density of the

moderator/coolant could reduce neutron absorption significantly while reducing

moderation only slightly, making the void coefficient positive. Also, light water is

actually a somewhat stronger moderator of neutrons than heavy water, though heavy

water neutron absorption is much lower. Because of these two facts, light water reactors

have a relatively small moderator volume and therefore have compact cores. One next

generation design, the supercritical water reactor, is even less moderated. A less

moderated neutron energy spectrum does worsen the capture ratio for 235U and

especially 239Pu, meaning that more fissile nuclei fail to fission on neutron absorption

and instead capture the neutron to become a heavier non fissile isotope, wasting one or

more neutrons and increasing accumulation of heavy transuranic actinides, some of

which have long half-lives[6].

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1.4.2.3 Fuel PWR fuel bundle. This fuel bundle is from a pressurized water

reactor of the nuclear passenger and cargo hip NS Savannah. Designed and built by the

Babcock and Wilcox Company. After enrichment the uranium dioxide (UO2) powder is

fired in a high-temperature, sintering furnace to create hard, ceramic pellets of enriched

uranium dioxide. The cylindrical pellets are then clad in a corrosion-resistant zirconium

metal alloy (Zircaloy) which is backfilled with helium to aid heat conduction and detect

leakages. Zircaloy is chosen because of its mechanical properties and its low absorption

cross section. The finished fuel rods are grouped in fuel assemblies, called fuel bundles,

which are then used to build the core of the reactor. A typical PWR has fuel assemblies of

200 to 300 rods each, and a large reactor would have about 150250 such assemblies

with 80100 tons of uranium in all. Generally, the fuel bundles consist of fuel rods

bundled 14 14 to 17 17. A PWR produces on the order of 900 to 1,500 MWe. PWR

fuel bundles are about 4 meters in length.

Refueling for most commercial PWRs is on an 1824 month cycle.

Approximately one third of the core is replaced each refueling, though some more

modern refueling schemes may reduce refuel time to a few days and allow refueling to

occur on a shorter periodicity.

1.4.2.5 Control In PWRs reactor power can be viewed as following steam (turbine)

demand due to the reactivity feedback of the temperature change caused by increased or

decreased steam flow.Boron and control rods are used to maintain primary system

temperature at the desired point. In order to decrease power, the operator throttles shut

turbine inlet valves. This would result in less steam being drawn from the steam

generators. This results in the primary loop increasing in temperature. The higher

temperature causes the reactor to fission less and decrease in power. The operator could

then add boric acid and insert control rods to decrease temperature to the desired point.

Reactivity adjustment to maintain 100% power as the fuel is burned up in most

commercial PWRs is normally achieved by varying the concentration of boric acid

dissolved in the primary reactor coolant. Boron readily absorbs neutrons and increasing

or decreasing its concentration in the reactor coolant will therefore affect the neutron

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activity correspondingly. An entire control system involving high pressure pumps

(usually called the charging and letdown system) is required to remove water from the

high pressure primary loop and re-inject the water back in with differing concentrations

of boric acid. The reactor control rods, inserted through the reactor vessel head directly

into the fuel bundles, are moved for the following reasons:

To start up the reactor.

To shut down the reactor.

To accommodate short term transients such as changes to load on the turbine.

The control rods can also be used

To compensate fornuclear poisoninventory.

To compensate fornuclear fuel depletion.

but these effects are more usually accommodated by altering the primary coolant boric

acid concentration.

In contrast, BWRs have no boron in the reactor coolant and control the reactor power by

adjusting the reactor coolant flow rate.

1.4.3 Three stage programme for Indian nuclear power plant

The current share of nuclear energy in India is 3% but it has fillip recent year

and is poised to grow rapidly with advanced technology and application .The key to

capacity nuclear power sector critically depends upon development of reliable nuclear

power station, encompassing complete fuel breeding .So far India has developed

adequate core competence in all respect of nuclear power station , which is given by

three- stage nuclear programme of Dr Homi J. Bhabha. The first stage utilizes natural

uranium as fuel and heavy water as moderator in

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Fig. 1.8 Three stage nuclear power generation programme of India

Pressurized Heavy Water Reactor. Second stage are based on fast breeder reactors

(FBR) which are fuelled by plutonium obtained by reprocessing of spent fuel of thermal

reactors .Fast reactor produces more fissile material than what they consume and thus

multiplication of fissile inventory and enhancement of installed capacity .Third stage will

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based on Thorium U-233 cycle. Timely implementation of this stage is very crucial for

meeting increasing carbon free energy demand in country. As we have large resource of

Thorium resource it is our duty to design reactor which will implement effectively

thorium fuel cycle thus research at BARC developed design of advanced heavy water

reactor (AHWR)

One of greatest obstacles in this path is how to properly handle highly radioactive

waste, control cooling and radiation at Accident. Answer for these are safety designed

advanced reactor, closed cycle operation, reprocessing of nuclear fuel is must. It can be

obtained satisfactorily with this paper which enhances interest innovatively of nuclear

power generation technology, and here are some options such as advanced reactor design,

Hybrid system, Fusion driven power plant, Hydrogen formation fuel IV stage power

generation, etc.

1.5 Advantages 1.5.1PWR reactors are very stable due to their tendency to

produce less power as temperatures increase; this makes the reactor easier to operate

from a stability standpoint.

1.5.2. PWR turbine cycle loop is separate from the primary loop, so the water in

the secondary loop is not contaminated by radioactive materials.

1.5.3 PWRs can passively scram the reactor in the event that offsite power is lost.

Control rods are held by electromagnets and fall by gravity when current is lost.

1.5.4 Almost 0 emissions (doesn't emit green house gases). It produces electricity

without pollution into the atmosphere.

1.5.5 They can be sited almost anywhere unlike oil which is mostly imported.

1.5.6 A small amount of matter creates a large amount of energy... so there is little

fear that we will run out of it A lot of energy from a single power plant

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1.5.7 A truckload of Uranium is equivalent in energy to 10,000+ truckloads of coal.

(Assuming the Uranium is fully utilized)

1.5.8 New reactor types have been designed to make it physically impossible to melt

down. As the core gets hotter the reaction gets slower, hence a run-away reaction

leading to a melt-down is not possible.

1.5.9 Theoretical reactors (traveling wave) are proposed to completely eliminate any

long-lived nuclear waste created from the process.

1.5.10 Theoretical Thorium reactors have many of the benefits of Uranium reactors

while removing much of the risk for proliferation as it is impossible to get weapons-

grade nuclear materials from Thorium.

1.6 Disadvantages 1.6.1More expensive to build the plant .Proliferation

concerns. Breeder reactors yield products that could potentially be stolen and turned into

an atomic weapon.

1.6.2 Waste products dangerous and need to be carefully stored for long time. The spent

fuel is highly radioactive and has to be carefully stored for many years after use. This

adds to the costs.

1.6.3 Waste is now contained in state of the art sealed containers and does not otherwise

harm anyone. Nuclear power plants can be dangerous to its surroundings and employees.

There has been a single case where a plant has gone through a meltdown and as a result

left people dead and its surroundings destroyed.

1.6.4 The coolant water must be highly pressurized to remain liquid at high

temperatures. This requires high strength piping and a heavy pressure vessel and hence

increases construction costs. The higher pressure can increase the consequences of aloss

of coolant accident. Additional high pressure components such as reactor coolant pumps,

pressurizer, steam generators, etc. are also needed. This also increases the capital cost and

complexity of a PWR power plant.

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1.6.5 Natural uranium is only 0.7% uranium-235, the isotope necessary for thermal

reactors. This makes it necessary to enrich the uranium fuel, which increases the costs

of fuel production. Ifheavy water is used, it is possible to operate the reactor with

natural uranium, but the production of heavy water requires large amounts of energy

and is hence expensive.

1.6.6 Because water acts as a neutron moderator, it is not possible to build a fast

neutron reactor with a PWR design. A reduced moderation water reactor may

however achieve a breeding ratio greater than unity, though this reactor design has

disadvantages of its own.

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CHAPTER 2

ADVANCES IN NUCLEAR POWER PLANT

2.1 Generation IV Concepts

The U.S. Department of Energy participates in the Generation IV

International Forum (GIF), an association of ten nations that seek to develop a new

generation of commercial nuclear reactor designs before 2030. Criteria for

inclusion of a reactor design for consideration by the GIF group include:

1. Sustainable energy (extended fuel availability, positive environmental

impact)

2. Competitive energy (low costs, short construction times)

3. Safe and reliable systems (inherent safety features, public confidence in

nuclear energy safety)

4. Proliferation resistance (does not add unduly to unsecured nuclear material)

and physical protection (secure from terrorist attacks) [4].

2.1.1 Gas-cooled Fast Reactor (GFR) The GFR uses helium coolant

directly to a gas turbine generator to produce electricity. This parallels PBMR and

original GT-MHR designs. The primary difference from these designs is that the

GFR would be a "fast", or breeder reactor. One favored aspect of the design is that

it would minimize the production of many undesirable spent fuel waste streams.

The reference design size is targeted to be 288 MWe with a deployment target date

of 2025. In addition to producing electricity the design might be used as a processheat source in the production of hydrogen.

2.1.2 Lead-cooled Fast Reactor (LFR) So far, most breeder

reactors have used molten metal technologies for their coolants. Many FBRs have

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used molten sodium, a metal with which there is considerable experience but

which has sometimes been difficult to handle. The LFR uses molten lead or a lead-

bismuth alloy as its coolant. One design favored in the Generation IV would result

in long periods between refueling, 15-20 years. Similar designs have been

investigated in Russia. Target ranges for the reactor would be 50-150 MWe. That

would be rather small by historic nuclear standards, but might meet localized

market needs. Designs as large as 1200 MWe have been suggested. Initial targeted

deployment would be in 2025. Proposed designs would favor electricity

production though proponents consider the production of process heat at LFR units

as possible.

2.1.3 Molten salt reactor (MSR) The MSR involves a circulatingliquid of sodium, zirconium, and uranium fluorides as a reactor fuel. The MSR has

been presented as providing a comparatively thorough fuel burn, safe operation,

and proliferation resistance. The initial reference design would be 1000 MWe with

a deployment target date of 2025. The design could use a wide variety of fuel

cycles. Temperatures for electricity production would not be as hot as for some

other advanced reactors but some process heat potential exist. Versions of the

MSR have been around for some time but never were implemented for commercial

uses.

2.1.4 Sodium-cooled fast reactor (SFR) Sodium-cooled fast reactors

have been the most popular design for breeder reactors. Designs have been

proposed under the technological roadmap ranging from 150 to 1700 MWe.

Molten metal technology is no longer new but several early SFR prototypes had

difficulty obtaining sustained operation. The BN-600 in Russia has been regarded

as highly reliable. Design supporters believe that the SFR promises superior fuel

management characteristics. The target deployment date of 2015 reflects the

considerable research that the design has already received. Earlier prototypes have

already been built in France, Japan, Germany, the United Kingdom, Russia, and

the United States since as early as 1951. Initial deployment would probably focus

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on electricity due to comparatively low outlet temperatures for the design.

2.1.5 Supercritical-water-cooled reactor (SCWR) The SCWR

design is to be the next step in LWR development and has been proposed with

alternatives that evolve from the BWR and PWR. SCWRs would operate at higher

temperatures and thermal efficiencies than present LWRs. The reference plant

would be 1700 MWe, above the upper end of present LWR designs. The

deployment target date is 2025. Most research on the design has been in Japan.

Designers intend the SCWR to be much less expensive to build than today's LWR

units though some of the economies appear to be shared by units now undergoing

certification. Operating cost savings are also anticipated.

2.1.6 Very-high-temperature reactor (VHTR) The VHTR is an

evolution from the HTGR family of reactors but would operate at even higher

temperatures than designs now undergoing pre-certification. In contrast with the

GFR, the VHTR would not be a breeder reactor, thus it would produce less

potentially usable fuel than it consumes. In addition to generating electricity, the

design would provide process heat that could be used in industrial activities

including hydrogen production and desalinization. Electricity generation targets

have not yet been set. Deployment is targeted for 2020, earlier than all but one

Generation IV design. This reflects earlier experience with parallel designs.

Electricity generation would involve a heat exchanger rather than directprocessing

through a turbine.

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2.2 Waste management Built on the concepts of green supply chain

management (G-SCM), this report presents a multi-objective optimization

programming approach to address the issue of nuclear power generation. In this study,

a linear multi-objective optimization model is formulated to optimize the operations of

both the nuclear power generation and the corresponding induced-waste reverse

logistics. Factors such as the operational risks induced in both the power generation

and reverse logistics processes are considered, which speculates that nuclear power

may still dominate the world electricity resources in the 21st century as new nuclear

power technology being developed to mitigate the induced negative effects .

Despite the importance of nuclear power generation in addressing the issue of the

growing worldwide electricity demand, there appears to be an urgent need for

developing a green supply chain-based operational methodology to systematically

manage the entire lifecycle of the corresponding product, including the generated

nuclear power and induced wastes. By incorporating the coefficient of risk aversion

into the expected utility approach, the model proposed in Eeckhoudt et al. (2000)

appears promising to assess the external costs of nuclear fuel cycles imposed on society

and the environment under various operational conditions of nuclear power generation.

Similarly, considering the corresponding influencing factors and induced environmental

risks, Margulies (2004) proposed a risk-optimization programming approach to

determining the location of nuclear power generators. As further pointed out in Cowing

et al. (2004), the short-term trade offbetween productivity and safety often exist in the

operation of nuclear power generation.

In addition, Lee et al. (2000) proposed a Life Cycle Assessment (LCA) based

methodology to evaluate the environmental impact of nuclear power generation, where

a respective radiological impact assessment procedure is incorporated to enhance the

validity of the analytical results. The LCA-based approach provides a cradle-to-grave

environmental assessment framework serving to quantify the corresponding

environmental effects of nuclear power generation induced in its life cycle.

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2.2.1 System specification. Despite the complexity of technical features of

nuclear power generation, the performance of a typical nuclear-power supply chain is

mainly driven by nuclear fuel cycle, which is composed of five sequential phases (a)

mining and processing of raw materials, (b) nuclear fuel purchasing, (c) nuclear-

material transportation,(d) nuclear power generation, and (e) nuclear waste

management. They are briefly described in the following.

a) Mining and processing of raw materials The purpose of this phase is

to extract nuclear fuel from the corresponding raw materials, e.g., plutonium-239,

uranium-233 and uranium-235, where Canada and Australia are the two main supply-

source countries sharing about 55% of the global uranium production. Through

sophisticated processing procedures using specific technology, the raw materials are

refined such that their degrees of purity can reach from 0.7% up to 2-5%for the

production of nuclear fuels. Note that generally, the production planning of nuclear fuels

performed by the raw-material supplier is mainly determined by two factors: (1) the

contracting orders from demanders (i.e.,the countries generating nuclear power), and (2)

the lead time. Nowadays, the aforementioned two factors are regulated by the

International Atomic Energy Agency (IAEA) with the suggested values of eight years

and six months, respectively. Correspondingly, through long-term contracts followingthe regulations of IAEA, the nuclear fuel demanders should provide the 8-year

contracting demands to the corresponding nuclear fuel sup-pliers to ensure that the

periodic nuclear fuels can be supplied half a year before nuclear power generation.

b) Nuclear fuel purchasing In contrast with the previous phase dominated by

the nuclear fuel suppliers, the phase of nuclear fuel purchasing is mainly performed by

the nuclear power generators, determining the performance of a nuclear power supply

chain. Due to the limited number of nuclear fuel supply sources, the transnational

nuclear fuel purchasing maneuvers broadly exist around the world, forming unique few-

to-many nuclear-fuel distribution channels to satisfy the need of the worldwide 432

nuclear power generators. Under the regulations of IAEA in the global nuclear fuel

purchasing activities, the nuclear fuel demanders may implement diverse purchasing

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strategies, e.g. specific contracting with multiple supply sources, together with

respective fuel inventory control strategies in response to the uncertainty and complexity

in the process of transnational nuclear fuel purchasing. There-fore, it is inferred that

such sophisticated inbound raw-material logistics maneuvers driven by the nuclear fuel

demanders may greatly influence the performance of the nuclear power supply chain.

c) Nuclear-material transportation The transportation of nuclear materials

mainly involves the activities of shipping nuclear fuels from the co-responding suppliers

to power generators, and transporting the induced nuclear wastes through reprocessing

to final disposal. Since the nuclear materials are radioactive, falling under of the

classification systems recommended by the United Nations, the corresponding

transnational shipment activities, including labeling and packing, should be directed by

the regulations drawn by IAEA, and follow respective requirements of related

international organizations, e.g., the International Maritime Organization (IMO) in

shipment (Won- ham et al., 2000). In the surface transportation aspect, trucks and rails

are the two typical modes available for the nuclear-material transportation; however,

should also be subjected to local regulations for safety concerns.

d) Nuclear power generation The efficiency and reliability of nuclear

power generation rely mainly on the technologies of reactors, which are the core

devices, and can be briefly classified into four types: (1) water cooled reactors, (2) gas

cooled reactors, (3) liquid metal cooled reactors, and (4) fast breeder reactors. Details

about the distinctive features of these technologies and their relative advantages can be

readily found elsewhere (Kupitz, 1995; Ponomarev-Stepnoi, 1997). Through inputting

certain nuclear fuels, the nuclear reactors are in operation to generate the given units of

nuclear power accompanying nuclear wastes. Compared to the general product

manufacturing phase, the distinctive feature of nuclear power generation is that the

generated nuclear power at this phase is unable to be stored, and needs to be consumed

immediately after production. Meanwhile, most of the induced nuclear wastes are highly

radioactive, and need to be well managed through the respective reverse logistics-related

procedures to alleviate the corresponding environmental and ecological impacts.

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Accordingly, both the power generation scheduling and logistics related maneuvers with

respect to nuclear fuels as well as induced wastes should be efficiently coordinated in

this phase for nuclear-power G-SCM.

e) Induced nuclear waste management Apparently, nuclear waste management is

essential in promoting the nuclear power generation mode. Briefly, the induced nuclear

wastes can be divided into two categories

Operating wastes (i.e., wastes accompanied with operation and decommissioning),

Spent nuclear fuels (i.e., wastes from the nuclear fuel cycle).

Therein, the management of spent nuclear fuels is the most critical issue in this phase

since the corresponding waste is highly radioactive, damaging both the environment and

ecology. Therefore, there are two potential alternatives to manage the spent nuclear fuels

Direct final disposal and Reprocessing-storage-final disposal, where the latter one is

broadly accepted around the world. In contrast, the operating wastes are low-level

radioactive, and can be handled through the procedures of high-pressure compaction,

solidification, transportation and final disposal.

G-SCM (Sheu- et al.,2005), a comprehensive conceptual generation. Here in,

seven primary layers are considered, including 1) nuclear fuel supplier, (2) nuclear

reactor, (3) electricity demand market, (4) reprocessing facility (for spent nuclear fuels),(5) interim storage facility, (6) low-level radioactive waste reprocessing facility, and (7)

final disposal. Among these, the first three layers are classified as the power supply chain

(sc for short) members, and the rest are the nuclear-waste reverse logistics chain (rc for

short) members. Each member defined here represents a specific logistics-related

function executed for G-SCM based nuclear power generation. It present study aims at

developing the prototype of a G-SCM based operational model for the multi-interval

nuclear power generation operation. Specific events such as the short-term outages and

the corresponding effects on the periodic power generation are not considered.

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Fig. 2.1. G-SCM based nuclear power generation[5].

In addition, to facilitate model formulation, four assumptions are postulated in the

following

a) The corresponding schedules for maintenance activities such as inspection,

repair and replacement occurring during the planning horizon are given.

b) The nuclear fuel cycle and the proportion of the amount of the induced

wastes relative to the nuclear fuels are known.

c) Facility locations and capacities associated with chain members of the

proposed G-SCM based nuclear power generation system are known.

d) The lead-time associated with each chain member either in the general

supply chain or in the reverse logistics chain is given.

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2.3 Fast Reactor Technology 2.3.1 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. Of the known and proven energy technologies,

only nuclear fission can provide the large quantities of energy required by industrial

societies in a sustainable and environmentally acceptable manner.

2.3.2 Natural uranium mined from the earth's crust is composed primarily of two

isotopes: 99.3% is U-238, and 0.7% is the fissile U-235. Nearly all current power

reactors are of the thermal neutron design, and their capability to extract the potential

energy in the uranium fuel is limited to less than 1% of that available. The remainder of

the potential energy is left unused in the spent fuel and in the uranium, depleted in U-

235, that remains from the process of enriching the natural uranium in the isotope U-235

for use in thermal reactors. With known fast reactor technology, this unutilized energy

can be harnessed, thereby extending by a hundred-fold the amount of energy extracted

from the same amount of mined uranium.

2.3.3 Fast reactors can convert U-238 into fissile material at rates faster than it is

consumed making it economically feasible to utilize ores with very low uranium

concentrations and potentially even uranium found in the oceans. A suitable technology

has already been proven on a small scale. Used fuel from thermal reactors and the

depleted uranium from the enrichment process can be utillized in fast reactors, and that

energy alone would be sufficient to supply the nations needs for several hundred years.

2.3.4 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.

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2.3.5 Although fast reactors do not eliminate the need for international proliferation

safeguards, they make the task easier by segregating and consuming the plutonium as it

is created. The use of onsite reprocessing makes illicit diversion from within the process

highly impractical. The combination of fast reactors and reprocessing is a promising

option for reasons of safety, resource utilization, and proliferation resistance.

2.3.6 Reaping the full benefits of fast reactor technology will take a decade or more for a

demonstration reactor, followed by build up of a fleet of operating power stations. For

now and in the intermediate-term future, the looming short-term energy shortage must

be met by building improved, proven thermal-reactor power plants. To assure longer-

term energy sustainability and security, the American Nuclear Society sees a need for

co-operative international efforts with the goal of building a fast reactor demonstration

unit with onsite reprocessing of spent fuel [7].

2.4 Challenges for nuclear power plant As we have discussed major

challenges such as waste disposal, nuclear proliferation, improving efficiency of nuclear

power plant along with safety( radioactive waste disposal.). Among these most burning

issue that has to face to nuclear power plant is accidental cooling challenge of power

plant (e.g. Seismic accident of Fukushima power plant laid it to complete melting of

plant, technology failed to control radiations, heat and cooling to save the power plant.)

2.4.1 Fast breeder reactors (FBR) are least safe Nuclear

safety has become top priority after Fukushima nuclear disaster in Japan. FBR will use

Indias most abound source of thorium. Conventional reactors make use of light or heavy

water for cooling. FBR make use of liquid sodium as coolant which is inherentlydangerous.Liquid sodium reacts with both air and water. Hence tiny leakage can cause

fire. Fukushima disaster has showed uncertainty may happen. If sudden things go happen

with FBR we cant have option of water and air for cooling, and sudden abundant sodium

cant make available. Thus it is least safe [8].

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2.4.2 Fusion driven nuclear power plant fusion offer possibility of

high density without generation of high level radioactive waste with long half life and

green house gases. It is process that powers the Sun. Two light nuclei of hydrogen

(deuterium, tritium) fuse together to form a heavier nuclei .In doing so large amount of

energy is released. But for that high very temperature, pressure has to achieve, which is

same like plasma. It is thus challenging to gain maximum energy by fusion power plant.

It has to maintain plasma conditions by holding it with powerful electromagnetic forces.

2.5 Conclusion

Despite challenges within steps of the nuclear fuel cycle, harnessing energy from

nuclear fission provides a promising and realistic source of power. The challenges that do

exist can be solved with improvements in technology and with global co-operation. The

sustainability of uranium mining activities will improve if countries implement more

environmentally conscious mining strategies. Positive communication among countries

and international governance of enrichment facilities by an impartial organization will

lead to the production of only low-enriched uranium.

Using plutonium/thorium based fuel makes the existence of radioactive material

safer as they cannot be used for weapons-grade material. Reprocessing and transmutation

will lessen the amount of the dangerous waste produced. Additionally, looking into

strategies will help the United States to deal with the amount of nuclear waste produced.

Lastly, in the India, making the public more educated and addressing the public

concerns will do much to increase public support for nuclear power, a necessary step in

Indias perusal of nuclear energy as a power source.

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References

1. Dr. R. Yadav., et al. Steam and gas turbines and power plant

engineering, Central publishing house, Allahabad, 7th Ed., 2004.

2. Mazen M. Abu-Khader., Recent advances in nuclear energy: A review,

51 (2009) 225-235.

3. P. K. Dey, and N.K.Bansal., Spent fuel reprocessing, vital link in Indian

nuclear power program, 236 (2006) 723-729.

4. John Sackett. Future of nuclear energy, 71 (2001) 197-204.

5. W.L.Wilkinson., A review; Challenges to nuclear power in UK,

84(A4):261-264 IChemE.

6. Jiuh-Biing Sheu. Green supply chain management, reverse logistics and

nuclear power generation, 44 (2008) 1946.

7. Fast Reactor Technology: A Path to Long-Term Energy Sustainability,

Position Statement. 74, Nov.2005,(708) 352 6611.

8. Dr. Swaminathans Aiyar., Fast breeder reactors are least safe, article in

Times of India, April 12, 2011.

.

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