A

REPORT ON

SUMMER TRANING

AT

“ELECTRIC POWER GENERATION IN RAJASTHAN

ATOMIC POWER STATION, RAWATBHATA”

SESSION-(2012-2013)

Department of Electrical Engineering

BALDEV RAM MIRDHA INSTITUTE OF TECHNOLOGY ,

JAIPUR

SUBMITTED TO: SUBMITTED BY:

MOHIT KHAROL RAKESH KUMAWAT

BMIT COLLEGE, JAIPUR B.TECH, 4th YR. (7th SEM.), EE

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PREFACE

I Rakesh Kumawat student of fourth year of Electrical Engineering have

completed practical training at Rajasthan Atomic Power Station (RAPS) for 30

days from 21/05/12 to 19/06/12.

Being an engineering student, the training at Rajasthan Atomic Power Station

(RAPS) has been particularly beneficial for me. I saw various procedures,

processes and equipments used in production of electricity by nuclear power,

which were studied in books, and thus helped me in understanding of power

generation and distribution concepts of electrical power.

Rajasthan Atomic Power Station, a constituent of board of Nuclear Power

Corporation Of India Limited is a very large plant & is very difficult to acquire

complete knowledge about it in a short span. I have tried to get acquainted with

overall plant functioning and main concepts involved therein.

RAKESH KUMAWAT

Electrical Engineering

B.M.I.T.,JAIPUR.

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ACKNOWLEDGEMENT

I am highly indebted and owe a sense of gratitude towards Mr.R.K.Sharma

Training Superintendent for giving me opportunity to impart training at Nuclear

Training Centre of RAJASTHAN ATOMIC POWER STATION under the

guidance of eminent professionals. It was highly educative and interactive to

take training at such a prestigious organization.

My sincere gratitude and thanks to Mr. R.C. Purohit , Senior Training Officer

and Training Co-ordinator, for providing me opportunity to complete my

training work at NTC.

I am also thankful to all those who helped me directly or indirectly through their

invaluable guidance and inspiration for successful completion of this training.

RAKESH KUMAWAT

B.M.I.T.,JAIPUR.

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TABLE OF CHAPTER

S. NO. CHAPTER PAGE NO.

1. Introduction 5

2. Rajasthan Atomic Power Station 6

3. Nuclear Reactor Technology 7

4. Indian Nuclear Power 21

5. Cataloging of Nuclear Reactors 23

6. Radioactive Waste Management 28

7. Safety 30

8. RAPPCOF 33

9. Fire Section 35

10. Environmental survey laboratory 36

11. Future of the Industry 37

12. View of different stations 37

13. Conclusion 40

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INTRODUCTION

Nuclear power is any nuclear technology designed to extract usable energy from

atomic nuclei via controlled nuclear reactions. The only method in use today is

through nuclear fission, though other methods might one day include nuclear

fusion and radioactive decay . All reactors heat water to produce steam, which

is then converted into mechanical work for the purpose of generating

electricity . In 2007, 14% of the world's electricity came from nuclear power.

More than 150 nuclear-powered naval vessels have been built, and a few

radioisotope rockets have been produced.

A nuclear reactor is a device in which nuclear chain reactions are initiated,

controlled, and sustained at a steady rate, as opposed to a nuclear bomb, in

which the chain reaction occurs in a fraction of a second and is uncontrolled

causing an explosion.

The most significant use of nuclear reactors is as an energy source for the

generation of electrical power and for the power in some ships .. This is usually

accomplished by methods that involve using heat from the nuclear reaction to

power steam turbines.

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RAJASTHAN ATOMIC POWER STATION

UNIT-1&2

Fig.-1 Unit-3&4

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Rawatbhata remote town in Chittorgarh district about 64 KMs, from Kota, an

industrial city of Rajasthan. The land selected is in between Rana Pratap Sagar

Dam &Gandhi Sagar Dam at the right bank of Chambal River. The water from

the reservoir of the Rana Pratap Sagar Dam serves the requirements of the

Nuclear Power Plants. There are four PHWR units of 100,200,220 MWe and

two units of 235 MW newly constructed which feed the Northern Grid as abase

load station. . For employees various colonies are constructed with all the

domestic facilities.

NUCLEAR REACTOR TECHNOLOGY

Just as many conventional thermal power stations generate electricity by

harnessing the thermal energy released from burning fossil fuels, nuclear power

plants convert the energy released from the nucleus of an atom, typically via

nuclear fission.

When a relatively large fissile atomic nucleus (usually uranium-235 or

plutonium-239) absorbs a neutron, a fission of the atom often results. Fission

splits the atom into two or more smaller nuclei with kinetic energy (known as

fission products) and also releases gamma radiation and free neutrons. A portion

of these neutrons may later be absorbed by other fissile atoms and create more

fissions, which release more neutrons, and so on.

This nuclear chain reaction can be controlled by using neutron poisons and

neutron moderators to change the portion of neutrons that will go on to cause

more fissions. Nuclear reactors generally have automatic and manual systems to

shut the fission reaction down if unsafe conditions are detected.

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A cooling system removes heat from the reactor core and transports it to another

area of the plant, where the thermal energy can be harnessed to produce

electricity or to do other useful work. Typically the hot coolant will be used as a

heat source for a boiler, and the pressurized steam from that boiler will power

one or more steam turbine driven electrical generators.

There are many different reactor designs, utilizing different fuels and coolants

and incorporating different control schemes. Some of these designs have been

engineered to meet a specific need. Reactors for nuclear submarines and large

naval ships, for example, commonly use highly enriched uranium as a fuel. This

fuel choice increases the reactor's power density and extends the usable life of

the nuclear fuel load, but is more expensive and a greater risk to nuclear

proliferation than some of the other nuclear fuels.

A number of new designs for nuclear power generation, collectively known as

the Generation IV reactors, are the subject of active research and may be used

for practical power generation in the future. Many of these new designs

specifically attempt to make fission reactors cleaner, safer and less of a risk to

the proliferation of nuclear weapons. Fusion reactors, which may be viable in

the future, diminish or eliminate many of the risks associated with nuclear

fission

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NULLEAR FISSION PROCESS

A complete chain reaction of nuclear fission is as shown in fig.

cause more fissions. In nuclear engineering, a neutron moderator is a medium

which reduces the velocity of fast neutrons, thereby turning them into thermal

neutrons capable of sustaining a nuclear chain reaction involving uranium-235.

Commonly used moderators include regular (light) water (75% of the

world's reactors), solid graphite (20% of reactors) and heavy water (5% of

reactors). Beryllium has also been used in some experimental types, and

hydrocarbons have been suggested as another possibility. Increasing or

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decreasing the rate of fission will also increase or decrease the energy output

of the reactor.

CALANDRIA

It is the heart of reactor and contains fuel and moderator; it is made of

Austenitic Stainless Steel. It contains 306 horizontal calandria tubes made

form Nickel- free- Zically-2. It also contains a special tube, which has 12

fuel bundles making a total of 3672 fuel bundles. It also has 6 openings at

the top through which pass the reactivity control mechanism assemblies. In

the middle it has piping connection for moderator outlet & inlet. The entire

assembly is supported from calandria vault roof.

Heat Generation

The reactor core generates heat in a number of ways:

The kinetic energy of fission products is converted to thermal energy

when these Some of the gamma rays produced during fission are

absorbed by the reactor in the form of heat.

Heat produced by the radioactive decay of fission products and materials

that have been activated by neutron absorption. This decay heat source

will remain for some time even after the reactor is shutdown.

The heat power generated by the nuclear reaction is 1,000,000 times that of the

equal mass of coal.

TURBINE

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Turbine is tandem compound machine directly coupled to electrical

generator. A turbine generally consists of low- pressure cylinder (double

flow for 500 MW units).

Turbine has a maximum continuous & economic rating of 229MW. Turbine

is the horizontal tandem compound re-heating impulse type running at

3000RPM with special provision for the extraction of moisture. A steam

turbine converts heat energy of steam into mechanical energy and drives the

generator. It uses the principle that the steam when issuing from a small

opening attains a high velocity. This velocity attained during expansion

depends on the initial and final heat content of steam. The difference

between initial & final heat content represents the heat energy converted into

mechanical energy.

STEAM GENERATORS

The boiler assemblies contain 10-u shaped shell & tube heat exchangers ,

connected in parallel. The hot coolant inlet channel and returning cold-water

channel are welded, the shell material is carbon steel & tube material is

Monel. Each heat exchangers has 195 tubes approximately 42 ft. long 4.5”

dia. 049 thou thick the design pressure on the heavy water side of the boiler

is 1350 psig at 5700 f.

COOLING

A cooling source - often water but sometimes a liquid metal - is circulated past

the reactor core to absorb the heat that it generates. The heat is carried away

from the reactor and is then used to generate steam. Most reactor systems

employ a cooling system that is physically separate from the water that will be

boiled to produce pressurized steam for the turbines, like the pressurized water

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reactor. But in some reactors the water for the steam turbines is boiled directly

by the reactor core, for example the boiling water reactor.

FUEL

The use of natural uranium dioxide fuel with its it s low content of fissile

material (0.72% u-235) precludes the Possibility of a reactivity accident

during fuel handling or storage. Also, in the core there would no significant

Increase in the reactivity, in the ever of any mishaps causing redistribution of

the fuel by lattice distortion or otherwise.

The thermal characteristics namely the low thermal conductivity and high

specific heat oh UO2, permit almost all the heat generated in a fast power

transient to be initially absorbed in the fuel. Furthermore, high melting point

of UO2 permits several full power seconds of heat to be safely absorbed that

contained at normal power.

Most of the fission products remain bound in the UO2 matrix and may get

released slowly only at temperatures considerably higher than the normal

operating temperatures. Also on the account of the uranium dioxide being

chemically inert to the water coolant medium, the defected fuel releases

limited amount of radioactivity to the primary coolant system.

The use of 12 short length fuel bundles per channels in a PHWR, rather than

full- length elements covering the whole length of the core, subdivides the

escapable radioactive facility in PHWR has also the singular advantage of

allowing the defected fuel to be replaced by fresh fuel at any time.

The thin zircoloy-2/4 cladding used in fuel elements is designed to collapse

under coolant pressure on to the pellets. This feature permits high pellet-

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clad gap conductance resulting in lower fuel temperature and consequently

lower fission gas release from the UO2 matrix into pellet- clad gap.

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FUEL DESIGN

Fuel assemblies in the reactor are short length (half meter long) fuel bundles.

Twelve of such bundles are located in each fuel channel. The basic fuel

material is in the form of natural uranium dioxide a pellet, sheathed & sealed

in thin Zircaloy tubes. Welding them to end plates to form fuel bundles

assembles these tubes.

FUEL HANDLING

On – power fuelling is a feature of all PHWRs, which have very low excess

reactivity. In this type of reactor, refueling to compensate for fuel depletion&

for over all flux shaping to give optimum power distribution is carried out

with help of 2 fueling machines, which work in conjunction with each other

on the opposite ends of a channel. One of the machines is used to fuel the

channel while the other one accepts the fuel bundles.

In addition, the fueling machines facilitate removal of failed fuel bundles.

Each fueling machine is mounted on a bridge & column assembly. Various

mechanisms provided along tri-directional movement (X, Y&Z Direction) of

fueling machine head and make it mechanisms have been provided which

enables clamping of fueling machine head to the end fitting, opening &

closing of the respective seal plugs, shield plugs & perform various fuelling

operations i.e. receiving new fuel in the magazine from fuel transfer system,

sending spent fuel From magazine to shuttle transfer station, from shuttle

transfer station to inspection bay & from inspection bay to Spent fuel storage

bay.

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MODERATOR SYSTEM

The heavy water moderator is circulated through the calandria by aid of a

low temperature & low- pressure moderator system. This system circulates

the moderator through two heat exchanges, which remove heat dissipated by

high- energy neutrons during the process of moderation. The cooled

moderator is returned to the calandria via. Moderator inlet nozzles. The high

chemical purity and low radioactivity level of the moderators are maintained

through moderator purification system. The purification systems consists of

stainless steel ion –exchange hoppers, eight numbers in 220MW contains

nuclear grade, mixed ion- exchange Resin (80% anion &20% cat ion resins).

The purification is also utilized for removable of chemical shim, boron to

affect start- up of reactor helium is used as a cover- gas over the heavy water

in calandria. The concentration deuterium in this cover –gas is control led by

circulating it using a sealed blower and passing through the recombination

containing catalyst alumina- coated with 0.3% palladium.

Primary Heat Transport (PHT) System

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Primary Heat Transport (PHT) System

The system, which circulates pressurized coolant through the fuel channels

to remove the heat generated in fuel, referred as Primary Heat Transport

System. The major components of this system are the reactor fuel channels,

feeders, two inlet headers, two reactor outlet headers, four pumps &

interconnecting pipe & valves. The headers steam generators & pumps are

located above the reactor and are arranged in two symmetrical banks at

either end of the reactor. The headers are connected to fuel channels through

individual feeder pipes. Figure 6 depicts schematically the relative layout of

major equipment in one bank of the PHT system. The coolant circulation is

mentioned at all times during reactor operation, shutdown& maintenance.

REACTIVITY CONTROL

The power output of the reactor is controlled by controlling how many neutrons

are able to create more fissions.

Control rods that are made of a nuclear poison are used to absorb neutrons.

Absorbing more neutrons in a control rod means that there are fewer neutrons

available to cause fission, so pushing the control rod deeper into the reactor will

reduce its power output, and extracting the control rod will increase it.

In some reactors, the coolant also acts as a neutron moderator. A moderator

increases the power of the reactor by causing the fast neutrons that are released

from fission to lose energy and become thermal neutrons. Thermal neutrons are

more likely than fast neutrons to cause fission, so more neutron moderation

means more power output from the reactors. If the coolant is a moderator, then

temperature changes can affect the density of the coolant/moderator and

therefore change power output. A higher temperature coolant would be less

dense, and therefore a less effective moderator.

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In other reactors the coolant acts as a poison by absorbing neutrons in the same

way that the control rods do. In these reactors power output can be increased by

heating the coolant, which makes it a less dense poison. Nuclear reactors

generally have automatic and manual systems to insert large amounts of poison

(boron) into the reactor to shut the fission reaction down if unsafe conditions are

detected.

ELECTRICAL POWER GENERATION

The energy released in the fission process generates heat, some of which can be

converted into usable energy. A common method of harnessing this thermal

energy is to use it to boil water to produce pressurized steam which will then

drive a steam turbine that generates electricity.

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REACTOR

The reactor is used to convert nuclear energy into heat. While a reactor could be

one in which heat is produced by fusion or radioactive decay.

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INDIAN NUCLEAR POWER

The Headquarters of Indian Nuclear Power Projects are located at Mumbai

known as the Department of Atomic Energy (DAE) which covers all the aspects

of R&D and power production. It is at Bhabha Atomic Research Centre where

all the research works regarding the new technologies and nuclear science.

Other than the power production plants there are various other

institutions that come under DAE like, Nuclear Fuel Compels (NFC) at

Hyderabad, Mines at Jadugura, and Centre for Advance Technology, Indore etc.

The first nuclear power plant was constructed at Tarapur in 1969. It

was a Boiling Water Reactor. The purpose of this reactor was to give the ground

for development of Pressurized Heavy Water Reactors (PHWRs). The two

units’ setup on turnkey basis by G.E., America is still working successfully.

The list of proposed sites for (PHWR) in India-

KAPP3&4 740X2

RAPP7&8 740X2

Jetpur(Maharastra) 740X2

The list of various Nuclear Power Plants in India is as follows:-

Station Rated Capacity

(MW)

Year of Criticality

TAPS-1&2 2 x 160 1969

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RAPS-1 100 1972 (S/D)

RAPS-2 200 1980

RAPS-3 235 1999

RAPS-4 235 2000

RAPS-5 235 Project under construction

RAPS-6 235 Project under construction

MAPS-1 220 1983

MAPS-2 220 1985

NAPS-1 220 1989

NAPS-2 220 1991

KAPP-1 220 1992

KAPP-2 220 1993

KAIGA-1 235 1996

KAIGA-2 235 1996

KAIGA-3 235 Project under construction

KAIGA-4 235 Project under construction

TAPS-3 540 2006

TAPP-4 540 2005

MADRAS 500 F/B reactor Project under construction

Kk project 1 1000 Light water reactor under construction

Kk project 2 1000 Light water reactor under construction

CATALOGING OF NUCLEAR REACTORS

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CLASSIFICATION OF REACTOR ON BASIS OF NEUTRON ENERGY:

Each fission process produces 2.5 new neutrons and, at least one of these must

produce a further fission for a chain reaction to be maintained. So for every 100

neutrons, produced in one neutron generation, at least 40 must cause further

fissions so as to produce 40 x 2.5 or 100 neutrons in the next generation. Now

the neutrons produced at fission are fast neutrons with an average energy of 2

MeV. If the fissions occur in natural uranium fuel, 99.3% of the nuclei are U-

238 is solitary responsible for the fission with neutrons having energies greater

than 1.2 MeV, therefore only half the fission neutrons can cause U-238 fissions.

So out of the 100 neutrons produced at fission, only 50 can cause U-238

fissions.

The inelastic scattering cross-section of U-238 is 10 times greater than the

fission cross-section at these neutron energies. So, out of these 50 neutrons 5

will be able to cause fission and remaining 45 will be scattered and lose so

much energy that they can no longer cause U-238 fission. The fast fission cross

section in U-235 is only 1.44 barns and U-235 fast fissions can be ignored with

so little U-235 in natural uranium. Therefore, out of the 100 fast neutrons

produced at fission only 5 will cause further fissions and produce 5 x 2.5 new

neutrons. Thus even if leakage and radioactive capture are ignored the chain

reaction can not be maintained by fast neutrons in natural uranium. One of two

alternatives is available which lead to a power reactor classification as follows:

FAST BREEDER REACTORS

The U-235 content of the fuel can be increased, i.e., the fuel is highly enriched

in U-235 with a substantial decrease in U-238. The U-235 fast fissions are thus,

considerably increased in a fast reactor. Some reduction in neutron energy does

occur due to inelastic collisions of neutrons with nuclei of the fuel and structural

material but most of the fissions are caused by neutrons of energies greater than

0.1Mev.The mass of U-235 required for the reactor to be critical varies with a

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mount of U-235 enrichment. In all cases the critical mass of fissile material

required increases rapidly below 15% to 20% U-235 enrichment. To avoid large

fuel inventories a fast reactor, would require fuel containing at least 20% U-235

by volume. Incidentally the critical mass of U-235 in a fast reactor is

considerably greater than in a thermal reactor with the same fuel composition.

The highly enriched fuel, absence of moderator results in a small core.

Therefore, fast reactors have high power density cores. The average power

density in a (FBR) is 500 MW/m3 compared with 100 MW/ m3 for a (PWR). It

is therefore essential that a heat transport fluid with good thermal properties be

used. The choice is also limited to a non-moderating fluid & liquid metals seem

to satisfy both requirements. The capture cross-sections of most elements for

fast neutrons are small & since there is a relatively large mass of U-235 in the

reactor, the macroscopic capture cross-sections of structural material and fission

products are small compared with the macroscopic fission cross-section of the

U235.Consequently there is more flexibility in the choice of materials and

stainless steel can be used instead of aluminum or zirconium. Fission product

poisoning is not significant as that temperature coefficient of reactivity is low;

the excess reactivity required in a fast reactor is small.

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THERMAL REACTORS

Since a chain reaction can not be maintained with fast neutrons without

considerable enrichment, the alternative is to reduce the neutron energy until the

fission cross-section of U-235 is sufficiently increased. If the neutrons are

reduced to thermal energies, the U-235 fission cross-section is 580 barns

whereas the radioactive capture cross-section is 106 barns. Thus, even allowing

for the low percentage of U-235 in natural uranium, the thermal neutron fission

cross-section in natural uranium is 4.2 barns whereas the radioactive capture

cross-section is 3.5 barns. Thus, for every 77 neutrons captured in natural

uranium about 40 will cause fission and produce 40 x 2.5 or 100 new neutrons.

For 77 neutrons out of every 100 to be captured, fewer than 23 neutrons can be

lost by escape or radioactive reaction could be sustained. In thermal reactors the

fission neutrons are thermalized by slowing them down in a moderator. Most of

the power reactors in existence are thermal reactors.

TYPES OF THERMAL REACTORS:

Previously, reactors were classified on the basis of neutron energy and the

various advantages and disadvantages of fast and thermal systems were

enumerated. It was mentioned that most of the reactor systems, at present in

operation, are thermal reactors. Thermal reactors will now be classified further

on the basis of core structure, the moderator used and the heat transport system

used. Some reference will be made to the advantages and disadvantages of each

type, but some of these considerations will be discussed later when moderator

and heat transport system properties are discussed.

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TYPES OF HETROGENEOUS REACTORS:

Classification of heterogeneous reactors may be based on the type of moderator

used or on the heat transport system employed. The basic requirements &

properties of moderators & heat transport systems will be discussed at length

later. It is sufficient, for the moment to list the moderators and heat transport

fluids in general use.

The moderator may be:

Light water

Heavy water

Graphite

Organic liquids

The heat transport system may be:

Pressurized light water

Pressurized heavy water

Boiling light water

Boiling heavy water

Gases such as CO2 or helium

Liquid metals

Steam or fog

Organic liquids

HEAVY WATER MODERATED REACTORS

These have much lower neutron capture cross section than both light water and

graphite. The principal advantage of using heavy water as a moderator is,

therefore, the neutron economy that can be achieved with it. The thermal

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utilization factor is increased because of lower neutron capture in the moderator.

Neutron economy is so much improved that not only can natural uranium fuel

be used, but that this fuel can be used in oxide/carbide form. Thus, there is no

longer need of enrichment plant. In addition oxide or carbide fuel improve the

fuel integrity & the fuel in less susceptible to distortion.

GRAPHITE MODERATED REACTORS

With a graphite moderator, a liquid or a gas must be used as the coolant.

Although there are water cooled graphite-moderated reactors, e.g., the Soviet

Union’s RBMK series of power stations, of which Chernobyl is one, only gas

cooled reactors will be referred to here. Whilst the United States and Canada

pioneered, respectively, the light and heavy water moderated designs, France

and United Kingdom undertook the early development of the graphite

moderated reactor, selecting carbon dioxide as the coolant because of its relative

chemical inertness and low neutron activation. France abandoned this approach

in favor of an extensive PWR programme. The UK continued to be heavily

committed to gas cooled reactors in the form, initially, of magnox and

subsequently the advanced gas cooled reactor.

PRESSURIZED HEAVY WATER REACTOR (PHWR):

PHWRs have established over the years a record for dependability, with load

factors in excess of 90% over extended periods. In the PHWR, the heavy water

moderator is contained in a large stainless steel tank (calandria) through which

runs several hundred horizontal zircaloy calandria tubes. The D2O moderator is

maintained at atmospheric pressure and a temperature of about 70°C.

Concentric with the calandria tube, but separated by a carbon dioxide filled

annulus which minimizes heat transfer from fuel to the moderator, is the

zircaloy pressure tube containing the natural UO2 fuel assemblies and the heavy

water coolant at a pressure of about 80 kg/cm² and a temperature of about

300°C.

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The term pressurized refers to the pressurized D2O coolant which

flows in opposite directions in adjacent tubes and passes its heat to the

secondary coolant via the steam generators. System pressure is maintained by a

pressurizing one of the legs of a steam generator.

RADIOACTIVE WASTE MANAGEMENT

Operation of a nuclear facility like nuclear power station inevitably leads to the

production of low level radioactive wastes which are collected segregated to

select best processing method, and conditioned for either interim site storage or

for disposal. The design of facilities is such that the average public exposure

from radioactive materials at the exclusion boundary is a small fraction of the

recommended AERB limits.

SOLID RADIOACTIVE WASTE MANAGEMENT SYSTEM :

Solid radioactive waste in segregated into three general categories based on

contact dose.

Category -1 Waste: Largely originates

Protective clothing . Contaminated metal parts and miscellaneous items.As it

can contain no radioactivity. This waste will be collected in unshielded standard

drums.

Category-II & III Waste. : Filter cartridges and ion exchanges resins

Typically this waste has an unshielded radiation field greater than 1 R/hr. on

contact. These require additional shielding and greater precautions than for

category-I during transportation, handling and storage operation.

LIQUID RADIOACTIVE WASTE MANAGEMENT SYSTEM:

The Liquid Radioactive Waste Management System provides for collection,

storage, sampling and necessary treatment and dispersal of any liquid waste

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produced by the station. The system is designed to control the release of

radioactivity in the liquid effluent streams so that radiations dose to members of

the public is with in those stipulated by the regulatory board. This system

handles radioactive wastes that are carried in liquid streams from the laundry

active floor drains, decontamination center and chemical laboratories.

GAS RADIOACTIVE WASTE MANAGEMENT SYSTEM:

An extensive ventilation system collects potentially active exhaust air from such

areas as the Reactor Building, the storage area, the decontamination center and

the heavy water management area. The active and potentially active exhaust air

and gases are all routed to a gaseous effluent exhaust duct. This exhaust flow is

monitored for noble gases, tritium, iodine and active particulate before being

released. Facilities for filtration are provided. Signals from the iodine, wide

range beta-gamma and particulate monitors are recorded in the control center.

Tritium monitoring is carried out by laboratory analysis.

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SAFETY

INDUSTRIAL SAFETY

We mean that the measures adopted as a whole in industry to reduce accidents

to bare minimum.

Factors responsible for Safety:

Plant layout

Design of machinery

Safety Gadgets and equipments

Protective aids

Safety culture & Respect for Safety

Attitude of the management/ employer - Caution Boards

Display of Good practices about Safety

Safety meetings, Open discussion and other measures

Safety Manual

Enforcement

Unsafe Act & Unsafe conditions

Causes of Accidents:

Hazards are the risks and perils or dangers that contribute to accidents and

injuries.

"HAZARDS DO NOT CAUSE ACCIDENTS, PEOPLE DO"

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Kinds of Hazards:

Fire

Heat

Material Handling

Floors

Ladders

Tools

Machinery

Walking and Working surfaces

Process

Chemicals

Electricity

Unsafe Act

Unsafe Condition

RADIATION SAFETY

Radiation in Nuclear reactor is produced in following ways :

Directly in fission reaction

By decay of fission products

Following types of radiations are encountered:

Alpha radiation

Beta radiation

Gamma radiation

Neutron radiation

Out of the above types of radiations Alpha radiation is practically zero,

whereas Beta and Gamma radiation fields may be present almost everywhere

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inside the reactor building and in negligible amount even outside the reactor

building. Neutron radiations are mainly present inside the reactor vault. It is

worth noting that the secondary side of the plant i.e. feed water and steam

cycle etc. are completely separate from the nuclear systems and are therefore

not supposed to be and neither they are to carry any sort of radioactive

particle and therefore free of contamination and radiation. It is also wroth noting

that all radiations are emitted from the nucleus of every radioactive nuclide

which will always have a tendency to become stable by emitting radiations

through disintegration.

The following reaction shows the emissions of Alpha, Beta, Gamma and

Neutron.

92U238 2He4 92U234 + (alpha)

It has very low penetrating power and can be stopped by simple paper.

1H3 2He3 (18 KeV) +beta

It also does not have good penetrating power and in human skin it can penetrate

up to about half mm. It can be very easily shielded

92U235 + 0n1 92U236 Xe + Kr + 0n1 + gamma + Heat

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Following methodologies are used to control the exposure to the radiation and

therefore resistive of the radiation dose.

(1) Administrative Control

(2) Zoning Technique

(3) Design Control

(4) Operation Control

(5) Maintenance and House keeping

Exposure to any kind of radiation can be controlled by an individual by

following methods:

(1) Distance

(2) Shielding

(3) Decay (Time to Decay)

RAPPCOF (COBALT FACILITY)

Here, recovery of COABALT-60 SLUGS/PELLETS from the IRRADIATEDHere, recovery of COABALT-60 SLUGS/PELLETS from the IRRADIATED

ABSORBER RODS received from different Nuclear Power Plants.ABSORBER RODS received from different Nuclear Power Plants.

2727CoCo5959 + +00nn1 1 2727CoCo6060 + +γγ

Thermal Thermal 00nn11 activation X-section: 37 Barns activation X-section: 37 Barns

Sp. Activity of Carrier free Co Sp. Activity of Carrier free Co60 60 : 1128 Ci/g: 1128 Ci/g

Half Life: 5.27 year Half Life: 5.27 year

Radiations: Radiations:

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β :0.31 MeV max. β :0.31 MeV max.

::γγ : 1.17 MeV 100%: 1.17 MeV 100%

::γγ : 1.33 MeV 100%: 1.33 MeV 100%

§Thermal Energy/1000 Ci : 4 cal/s§Thermal Energy/1000 Ci : 4 cal/s

§Radiation field at 1 mtr from 1 Ci : 1.35 R/hr§Radiation field at 1 mtr from 1 Ci : 1.35 R/hr

SLUGS/PELLETS:SLUGS/PELLETS:

The facility is designed to handle about 1 Mega Curies of Co-60. In order to

meet the demand of high and medium specific activity Co-60 and also for the

fabrication of sources of various sizes and shapes, cobalt is irradiated in the

form of nickel coated pellets of 1 mm dia x1 mm ht for production of high

specific activity Co-60 (> 100 Ci/g) and in the form of aluminum clad slugs 6

mm dia x 25 mm ht for the production of specific activity between 30-100

Ci/g.

Recovery of Co-60 from Cobalt Adjusters:

The cobalt adjusters are brought to RAPPCOF from power stations in a specialThe cobalt adjusters are brought to RAPPCOF from power stations in a special

shielding flask. For complete recovery of cobalt activity, the followingshielding flask. For complete recovery of cobalt activity, the following

operations are carried out in a sequence:operations are carried out in a sequence:

1. Discharging of adjuster into pool1. Discharging of adjuster into pool

2. Dismantling of adjuster in pool2. Dismantling of adjuster in pool

3. Transportation of sub-assemblies from pool to Recovery Cell3. Transportation of sub-assemblies from pool to Recovery Cell

4. Cell door operation4. Cell door operation

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5. Recovery of slugs/pellet capsules from sub-assemblies5. Recovery of slugs/pellet capsules from sub-assemblies

6. Recovery of pellets6. Recovery of pellets

7. Preparation of transport pencils for slugs7. Preparation of transport pencils for slugs

8. Preparation of pellet capsules for transportation8. Preparation of pellet capsules for transportation

9. Measurement of activity9. Measurement of activity

10. Loading of cobalt in transport flask10. Loading of cobalt in transport flask

11. Transportation of cobalt shielding flask11. Transportation of cobalt shielding flask

FIRE SECTION

RAPS have one common fire section from unit 1-6. It is located at 3&4 unit

area .For fire production mainly three things are required

1)fuel for burning

2) oxygen to support fire and

3) the third one is temperature.

For fire extinguishing we remove any one out of these three things.

CLASSIFICATION OF FIRE

S.

N

O

.

CLASS OF

FIRE

SOURCE OF FIRE BEST EXTINGUISER

1. A wood, paper, ordinary combustibles Soda, acid, water

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2. B Oil,paints,grease,dasoline,disel,petrol Foam, co2

3. C Fire in gaseous substances(H2) Co2 gas

4. D Fire in chemicals, metals Co2, dry chemical

5. E Electrical fire Co2, dry chemical

FIRE DETECTORS-

a.) smoke detectors

b.) temperature detectors

ENVIRONMENTAL SURVEY LABORATORY

(1)OBJECTIVES OF E.S.L. LAB AT RAWATBHATA-

Measurements of concentration of radio nuclides in various environmental

matrices collected from the environment of rawatbhata nuclear site.

ATMOSPHERIC TERRESTRIAL AQUATIC

Air tritium Soil Water

Rain water Grass Silt

Sulphide Cereals Sedim

Air particulate Pulses Fish

Milk Weed

• Measurement of internal contamination due to gamma emitting radio

nuclides by whole body counting of RAPS radiation workers.

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• Measurement of direct radiation exposure using environmental thermo

luminescent dosimeters.

• Computation of radiation does to the public and demonstrate compliance

with applicable regulatory limits

FUTURE OF INDUSTRY:

The nuclear power programme in India up to year 2020 is based on

installation of a series of MWe & 500MWe pressurized heavy water

reactor (PHWR) UNITS. 1000MWe light water reactors (LWR) coming

two 5 year plans. The total installed capacity of nuclear generation would

increase UNITS & fast breeder reactors (FBR) units. NPCIL plans to

contribute about 10% of the total additional needs of power of about

10000MWe per year i.e. 1000 MWe per year .

VIEW OF DIFFERENT STATIONS

RAJASTHAN ATOMIC POWER STATION-1&2

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RAJASTHAN ATOMIC POWER STATION-3&4

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RAJASTHAN ATOMIC POWER PROJECT-5&6

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CONCLUSION

The practical training at R.A.P.S. has proved to be quite faithful. It proved an

opportunity for encounter with such huge components like 220MW generators,

turbines, transformers and switchyards etc. The way various units are linked and

the way working of whole plant is controlled make the students realize that

engineering is not just learning the structure description and working of various

machines, but the greater part is of planning, proper management.

It also provides an opportunity to learn technology used at proper place and

time can save a lot of labor for example almost all the controls are computerized

because in running condition no any person can enter in the reactor building.

But there are few factors that require special mention. Training is not carried

out into its tree spirit. It is recommended that there should be some practical

work specially meant for students where the presence of authorities should be

ensured. There should be strict monitoring of the performance of students and

system of grading be improved on the basis of the work done. However training

has proved to be quite faithful. It has allowed as an opportunity to get an

exposure of the practical implementation to theoretical fundamental.

Prepared by :

RAKESH KUMAWAT

B.Tech (Electrical Engineering)

B.M.I.T.

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