Contents

1.INTRODUCTION ................................................................................................................ 3

1.1 About BARC ............................................................................................................... 4

1.2 Founder ....................................................................................................................... 4

2. ATOMIC ENERGY IN INDIA ............................................................................................. 5

2.1 Future Perspectives ................................................................................................... 5

2.2 Strategy For Nuclear Energy ..................................................................................... 6

2.2.1 Indian Nuclear Power Generation ................................................................... 6

2.2.2 Reprocessing of Spent Fuel ……………………………………………………...7

3. EVOLUTION OF NUCLEAR REACTORS ........................................................................ 9

3.1 PHWR (Pressurised Heavy Water Reactor) ............................................................. 9

3.2 Fast Breeder Reactors ............................................................................................. 10

3.3 Boiling Water Reactor ............................................................................................. 11

3.3.1 Introduction........................................................................................................ 11

3.3.2 Overview of BWR ............................................................................................... 12

4. NUCLEAR FUELS .......................................................................................................... 13

4.1 Introduction .............................................................................................................. 13

4.2 Types of Nuclear Fuels ........................................................................................ 14

4.2.1 Oxide fuel ........................................................................................................... 14

4.2.2 Metal Fuel ........................................................................................................... 14

4.2.3 Ceramic Fuels .................................................................................................... 15

4.2.4 Common physical forms of nuclear fuel ......................................................... 16

5. VARIOUS PLANTS AT BARC TARAPUR ..................................................................... 17

5.1 PREFRE (Power Reactor Fuel Reprocessing) Plant ............................................. 17

5.2 SSSF (Solid Waste Storage Surveillance Facility) Plant ....................................... 17

5.3 AFFF (Advanced Fuel Fabrication Facility) ........................................................... 18

6. MECHANICAL WORKSHOP AT AFFF, BARC .............................................................. 19

6.1 INTRODUCTION........................................................................................................ 19

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6.2 WORKSHOP ............................................................................................................. 19

6.2.1 LATHE MACHINE ............................................................................................... 19

6.2.2 MILLING MACHINE ............................................................................................ 21

6.2.3 SHAPING MACHINE .......................................................................................... 22

6.2.4 Centreless Grinding machine ........................................................................... 23

6.2.5 GRINDING MACHINE ......................................................................................... 24

6.2.6 ROLLING MACHINE ........................................................................................... 26

6.2.7 POWER SHEARING MACHINE ......................................................................... 26

6.2.8 COOLANT ........................................................................................................... 27

6.2.9 TOOL .................................................................................................................. 28

7. WWEELLDDIINNGG ........................................................................................................................ 29

7.1 TIG WELDING (GTAW) ............................................................................................. 29

7.1.1 Metals that can be welded ................................................................................ 30

7.1.2 Technical chart for TIG welding ....................................................................... 30

7.1.3 Technical chart for cutting process for Argon gas ........................................ 30

7.2 SMAW (SHIELDING METAL ARC WELDING .......................................................... 31

7.2.1 Join quality and strength .................................................................................. 31

7.2.2 Metals commonly welded ................................................................................. 31

7.3 WELDING ELECTRODE SPECIFICATION .............................................................. 32

7.4 WELDING DEFECTS ................................................................................................ 33

7.4.1 Cracks ................................................................................................................ 33

7.4.2 Distortion ........................................................................................................... 33

7.4.3 Incomplete penetration ..................................................................................... 33

7.4.4 Slag conclusion ................................................................................................. 34

8. ENCLOSURE BOX PANEL ............................................................................................ 35

8.1 INTRODUCTION........................................................................................................ 35

8.2 PANEL ....................................................................................................................... 35

8.2.1 FRONT PANEL ................................................................................................... 35

8.2.2 BACK PANEL ..................................................................................................... 35

8.2.3 TOP PANEL ........................................................................................................ 36

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8.2.4 SIDE PANEL ....................................................................................................... 36

8.3 DESIGN CONCIDERATION ...................................................................................... 36

8.4 SELECTION OF MATERIAL ..................................................................................... 36

8.4.1 STAINLESS STEEL ............................................................................................ 36

8.5 FABRICATION OF PANEL ....................................................................................... 38

8.5.1 Welding inspection method .............................................................................. 39

8.5.2 Principle of the Die Penetrant test ................................................................... 39

8.5.3 LEAK TESTING OF PANELS ............................................................................. 40

9. CONCLUSION…………………………………………………………………………………..41

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1. INTRODUCTION

1.1 ABOUT BARC:

The Bhabha Atomic Research Centre (BARC) is India's premier nuclear research

facility based in Tarapur, Mumbai. BARC is a multi-disciplinary research Centre with

extensive infrastructure for advanced research and development covering the entire

spectrum of nuclear science, engineering and related areas.

BARC's core mandate is to sustain peaceful applications of nuclear energy,

primarily for power generation. It manages all facets of nuclear power generation, from

theoretical design of reactors, computerized modeling and simulation, risk analysis,

development and testing of new reactor fuel materials, etc. It also conducts research in

spent fuel processing, and safe disposal of nuclear waste. Its other research focus areas

are applications for isotopes in industries, medicine, agriculture, etc.

1.2 FOUNDER:

Dr. Homi Jehangir Bhabha was the visionary who

conceptulised the Indian Nuclear Programme and along

with a handful of Scientists initiated the nuclear science

research in India in March, 1944. He envisaged the vast

potential of nuclear energy and its possible successful

utilization in the field of power generation and allied areas.

Dr. Bhabha started working with the goal of achieving self-

reliance in the fields of nuclear science and engineering

and today‘s Department of Atomic Energy which is a

consortium of different and diversified fields of science

and engineering is the final outcome of the farsighted

planning of Dr. Bhabha. Thus, in his own words ―When

Nuclear Energy has been successfully applied for power

production in, say a couple of decades from now, India will not have to look abroad for its

experts but will find them ready at hand‖.

Dr. Homi Jehangir Bhabha, realizing the immense potential of nuclear energy as

a viable alternative source for electric power generation, launched the Indian Nuclear

Programme in March 1944. It was the farsightedness of Dr. Bhabha to start nuclear

research in India at a time following the discovery of nuclear fission phenomena by Otto

Hahn and Fritz Strassman and soon after Enrico Fermi etal from Chicago reporting the

feasibility of sustained nuclear chain reactions. At that time very little information was

available to the outside world about nuclear fission and sustained chain reactions and

nobody was willing to subscribe to the concept of power generation based on nuclear

energy.

5

2. ATOMIC ENERGY IN INDIA

2.1 FUTURE PERSPECTIVES:

Atomic Energy has got a definite and decisive role to perform in the Indian Power

Generation and supply sector. Being a developing country, a major share of India's

overall electricity requirements has to be from non-conventional sources as the

conventional sources has got limitations to meet the galloping needs. India has achieved

self-sufficiency in the Nuclear Science and Technology thanks to the pioneering efforts

initiated by Dr. Homi Bhabha who visualized the Indian Nuclear Program and since then

meticulously carried on by the dedicated scientists and engineers of DAE family.

An adequate and uninterrupted power generation is an intrinsic essentiality for

the overall development of any nation. In quantitative terms, the per capita consumption

of electric energy is regarded as an indicative parameter of the socio economic growth

rate of a nation.

The major contribution to India’s power production programme comes from:-

Coal based thermal power stations (105,437 MW in 2012, ~ 55.3% of total power

output)

Hydroelectric power generation (38,848 Mw in 2012, ~ 20.38 % of total Power

Output)

Nuclear power generation (4,780 Mw, ~2.5% of total Power Output)

Non - conventional sources (wind, tidal etc.)(22,233 MW, ~ 11.6% of total Power

Output)

Per capita power consumption in India is around 600 Kwh/yr., which is much

below the world average consumption of 2430 Kwh/yr. Thus, massive increase in the

power generation to match the world average consumption is needed in the coming

years to enhance the overall national growth rate.

Our conventional resources are far from being adequate to achieve any

ambitious target in terms of power generation. With the depleting coal deposits and the

limited potential of hydel power, the nation‘s future requirements of power could be met

by tapping nuclear and other non - conventional resources. There is a lot of potential in

non-conventional sources and this must be harnessed.

By their very nature, while other non-conventional sources are suitable for small-

decentralized applications, nuclear power stations are suitable for large central

generating stations.

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2.2 STRATEGY FOR NUCLEAR ENERGY:

India has consciously proceeded to explore the possibility of tapping nuclear

energy for the purpose of power generation and the Atomic Energy Act was framed and

implemented with the set objectives of using two naturally occurring elements Uranium

and Thorium having good potential to be utilized as nuclear fuel in Indian Nuclear Power

Reactors. The estimated natural deposits of these elements in India are:

Natural Uranium deposits - ~70,000 tonnes

Thorium deposits - ~ 3,60,000 tonnes

2.2.1 Indian Nuclear Power Generation: Envisages A Three Stage Programme:

STAGE 1 » Pressurised Heavy Water Reactor using

STAGE 2 » Fast Breeder Reactor

STAGE 3 » Breeder Reactor

STAGE 1 » Pressurised Heavy Water Reactor using

Natural UO2 as fuel matrix

Heavy water as moderator and coolant

Natural U isotopic composition is 0.7 % fissile U-235 and the rest is U-238. In

the reactor

The first two plants were of boiling water reactors based on imported

technology. Subsequent plants are of PHWR type through indigenous R&D

efforts. India achieved complete self- reliance in this technology and this stage

of the programme is in the industrial domain.

The future plan includes:

Setting up of VVER type plants based on Russian Technology is under

progress to augment power generation.

MOX fuel (Mixed oxide) is developed and introduced at Tarapur to conserve

fuel and to develop new fuel technology.

STAGE 2 » Fast Breeder Reactor

India‘s second stage of nuclear power generation envisages the use of Pu-239

obtained from the first stage reactor operation, as the fuel core in fast breeder reactors

(FBR). The main features of FBTR are:-

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Pu-239 serves as the main fissile element in the FBR

A blanket of U-238 surrounding the fuel core will undergo nuclear transmutation to

produce fresh Pu-239 as more and more Pu-239 is consumed during the

operation.

Besides a blanket of Th-232 around the FBR core also undergoes neutron

capture reactions leading to the formation of U-233. U-233 is the nuclear reactor

fuel for the third stage of India‘s Nuclear Power Programme.

It is technically feasible to produce sustained energy output of 420 GWe from

FBR.

Setting up Pu-239 fuelled fast Breeder Reactor of 500 MWe power generation is

in advanced stage of completion. Concurrently, it is proposed to use thorium-

based fuel, along with a small feed of plutonium-based fuel in Advanced Heavy

Water Reactors (AHWRs). The AHWRs are expected to shorten the period of

reaching the stage of large-scale thorium utilization.

STAGE 3 » Breeder Reactor

The third phase of India‘s Nuclear Power Generation programme is, breeder

reactors using U-233 fuel. India‘s vast thorium deposits permit design and operation of

U-233 fuelled breeder reactors.

U-233 is obtained from the nuclear transmutation of Th-232 used as a blanket in

the second phase Pu-239 fuelled FBR.

Besides, U-233 fuelled breeder reactors will have a Th-232 blanket around the U-

233 reactor core which will generate more U-233 as the reactor goes operational

thus resulting in the production of more and more U-233 fuel from the Th-232

blanket as more of the U-233 in the fuel core is consumed helping to sustain the

long term power generation fuel requirement.

These U-233/Th-232 based breeder reactors are under development and would

serve as the mainstay of the final thorium utilization stage of the Indian nuclear

programme. The currently known Indian thorium reserves amount to 358,000

GWe-yr of electrical energy and can easily meet the energy requirements during

the next century and beyond.

2.2.2 Reprocessing of Spent Fuel » By an Open Cycle or a Closed Cycle mode.

“Open cycle” refers to disposal of the entire waste after subjecting to proper

waste treatment. This Results in huge underutilization of the energy potential of

Uranium (~ 2 % is exploited)

8

“Closed cycle” refers to chemical separation of U-238 and Pu-239 and further

recycled while the other radioactive fission products were separated, sorted out

according to their half-lives and activity and appropriately disposed of with

minimum environmental disturbance.

Both the options are in practice.

As a part of long – term energy strategy, Japan and France has opted ―closed

cycle‖

India preferred a closed cycle mode in view of its phased expansion of nuclear

power generation extending through the second and third stages.

Indigenous technology for the reprocessing of the spent fuel as well as waste

management programme has been developed by India through its own

comprehensive R&D efforts and reprocessing plants were set up and are in

operation thereby attaining self - reliance in this domain.

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3. EVOLUTION OF NUCLEAR REACTORS

3.1 PHWR (PRESSURISED HEAVY WATER REACTOR):

India‘s first stage of Nuclear Programme was based on the PHWR Technology for the

following advantages.

Optimum utilization of the limited uranium resources

Higher Plutonium yield, for the second stage fuel

Availability of Indigenous Technology

The most significant feature of the PHWR design is

Multiple pressure tube configurations instead of a large pressure vessel.

The first two reactors (1x100 MWe & 1x200 MWe) were built at Rawatbhata near

Kota in Rajasthan with the Canadian collaboration and became operational in the

year 1973 & 1981.

Two units located at Kalpakkam near Madras built later were of the same design

but using indigenous technology and were dedicated to the nation in the year 1984

& 1986.

Subsequently, the Reactors at Narora offered first opportunity to our engineers to

evolve an indigenous design based on operating experience and other

requirements such as stringent safety norms and seismic design. 2x220 MWe

PHWR‘s at Narora was connected to the grid in 1991 and 1992.

2x220 MWe PHWR‘s at Kakrapar became operational in 1993 and 1995 followed

by the 2x220 MWe PHWR‘s built at Kaiga and 2x220 MWe PHWR‘s at Rajasthan

in the year 2000.

10

In 2007, 1x220 MWe PHWR unit at Kaiga became available followed by 1x220

MWe PHWR in 2011.

2x220 MWe units at Rajasthan became functional in the year 2011.

The design of 540 MWe PHWR is the next step in the process of evolution and the

first two units based on this design were built at Tarapur. The First Unit was

dedicated to the Nation in 2005 and the second in 2006 and both the units are

working well. Technology for the manufacture of various components and

equipment is now well established and has evolved through active collaboration

between the DAE and the industry. Several universities and national institutions

have also participated in the development of PHWR technology apart from in

house efforts in DAE. As we gain experience and master technology, performance

of our plants is improving.

700 MWe PHWR

3.2 FAST BREEDER REACTORS:

Fig.3.2: Fast Breeder reactor

India‘s first 40 MWt Fast Breeder Test Reactor (FBTR) attained criticality on 18

October, 1985.

India becomes the sixth nation having the technology to build and operate a

FBTR besides USA, UK, France, Japan and the then USSR.

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The unique features of Indian FBTR are:

Indigenously developed U-Pu carbide fuel rich in Pu

Design, development and fabrication of all machineries, peripheral units and

materials are by the Indian Scientists in close coordination with industry.

Status Initial operational problems sorted out and the reactor operates smoothly

at a steady power level of 10.5 Mwt- maximum possible powers output owing to its small

core.

Future plans based on the Design, setting up and operation of FBTR has

provided rich experience and immense information with liquid metal cooled Fast Breeder

Reactor Technology and also confidence to embark upon the design of a 500 MWe

prototype fast breeder reactor [PFBR], is in advanced stage of completion at Kalpakkam.

PFBR design requires

A detailed and complete understanding of thermal- hydraulics phenomena

Creep, creep-fatigue interaction, and buckling and fluid- structure interaction for

the design optimization and also for an assessment of structural integrity.

A large number of codes, in the disciplines of thermal- hydraulics and structural

mechanics have been developed.

The codes have been validated either through experimental data or through

international benchmark tests.

Engineering R&D

For fast breeder reactor programme through simulated experiments and

component development.

Experimental data for validating the analytical codes and performance evaluation

codes.

Facilities to carry out these experiments in air, water and sodium environment.

Expertise for modeling phenomena, special instrumentation for measuring flow

patterns, vibration etc. and interpretation of data.

Capability to set up high temperature sodium facilities and their safe operation.

3.3 BOILING WATER REACTOR :

3.3.1 Introduction: The boiling water reactor (BWR) is a type of light water nuclear reactor used

for the generation of electrical power. It is the second most common type of electricity-

generating nuclear reactor after the pressurized water reactor (PWR), also a type of

light water nuclear reactor.

12

Fig.3.3: Boiling Water Reactor

The main difference between a BWR and PWR is that in a BWR, the reactor core

heats water, which turns to steam and then drives a steam turbine. In a PWR, the reactor

core heats water, which does not boil. This hot water then exchanges heat with a lower

pressure water system, which turns to steam and drives the turbine. The BWR was

developed by the Idaho National Laboratory and General Electric in the mid-1950s. The

main present manufacturer is GE Hitachi Nuclear Energy, which specializes in the design

and construction of this type of reactor.

3.3.2 Overview of BWR:

The BWR uses demineralized water as a coolant and neutron moderator. Heat

is produced by nuclear fission in the reactor core, and this causes the cooling water to

boil, producing steam. The steam is directly used to drive a turbine, after which it is

cooled in a condenser and converted back to liquid water. This water is then returned

to the reactor core, completing the loop.

The cooling water is maintained at about 75 atm (7.6 MPa, 1000–1100 psi) so

that it boils in the core at about 285 °C (550 °F). In comparison, there is no significant

boiling allowed in a PWR (Pressurized Water Reactor) because of the high pressure

maintained in its primary loop—approximately 158 atm (16 MPa, 2300 psi). Prior to the

Fukushima I nuclear accidents, the core damage frequency of the reactor was

estimated to be between 10−4 and 10−7 (i.e., one core damage accident per every

10,000 to 10,000,000 reactor years).

13

4. NUCLEAR FUELS

4.1 INTRODUCTION:

Nuclear fuel is a material that can be 'consumed' by nuclear fission or fusion to

derive nuclear energy. Nuclear fuel can refer to the fuel itself, or to physical objects (for

example bundles composed of fuel rods) composed of the fuel material, mixed with

structural, neutron moderating, or neutron reflecting materials.

Most nuclear fuels contain heavy fissile elements that are capable of nuclear

fission. When these fuels are struck by neutrons, they are in turn capable of emitting

neutrons when they break apart. This makes possible a self-sustaining chain reaction

that releases energy with a controlled rate in a nuclear reactor or with a very rapid

uncontrolled rate in a nuclear weapon.

The most common fissile nuclear fuels are uranium-235 (235U) and plutonium-

239 (239Pu). The actions of mining, refining, purifying, using, and ultimately disposing of

nuclear fuel together make up the nuclear fuel cycle.

Not all types of nuclear fuels create power from nuclear fission. Plutonium-238

and some other elements are used to produce small amounts of nuclear power by

radioactive decay in radioisotope thermoelectric generators and other types of atomic

batteries. Also, light nuclides such as tritium (3H) can be used as fuel for nuclear fusion.

Nuclear fuel has the highest energy density of all practical fuel sources.

Fig.4.1: Nuclear Fuel Processing cycle

14

4.2 TYPES OF NUCLEAR FUELS:

4.2.1 Oxide fuel:

For fission reactors, the fuel (typically based on uranium) is usually based on

the metal oxide; the oxides are used rather than the metals themselves because the

oxide melting point is much higher than that of the metal and because it cannot burn,

being already in the oxidized state.

UOX:

Uranium dioxide is a black semiconductor solid. It can be made by reacting uranyl nitrate with a base (ammonia) to form a solid (ammonium uranate). It is heated (calcined) to form U3O8 that can then be converted by heating in an argon / hydrogen mixture (700 °C) to form UO2. The UO2 is then mixed with an organic binder and pressed into pellets, these pellets are then fired at a much higher temperature (in H2/Ar) to sinter the solid. The aim is to form a dense solid which has few pores. The thermal conductivity of uranium dioxide is very low compared with that of zirconium metal, and it goes down as the temperature goes up.

MOX:

Mixed oxide, or MOX fuel, is a blend of plutonium and natural or depleted

uranium which behaves similarly (though not identically) to the enriched uranium feed

for which most nuclear reactors were designed. MOX fuel is an alternative to low

enriched uranium (LEU) fuel used in the light water reactors which predominate

nuclear power generation.

Some concern has been expressed that used MOX cores will introduce new

disposal challenges, though MOX is itself a means to dispose of surplus plutonium by

transmutation.

4.2.2 Metal Fuel:

Metal fuels have the advantage of much higher heat conductivity than oxide

fuels but cannot survive equally high temperatures. Metal fuels have a long history of

use, stretching from the Clementine in 1946 to many test and research reactors. Metal

fuels have the potential for the highest fissile atom density. Metal fuels are normally

alloyed, but some metal fuels have been made with pure uranium metal. Uranium

alloys that have been used include uranium aluminum, uranium zirconium, uranium

silicon, uranium molybdenum, and uranium zirconium hydride. Any of the

aforementioned fuels can be made with plutonium and other actinides as part of a

closed nuclear fuel cycle. Metal fuels have been used in water reactors and liquid

metal fast breeder reactors.

15

TRIGA Fuel:

TRIGA fuel is used in TRIGA (Training, Research, Isotopes, General Atomics)

reactors. The TRIGA reactor uses uranium-zirconium-hydride (UZrH) fuel, which has a

prompt negative temperature coefficient, meaning that as the temperature of the core

increases, the reactivity decreases—so it is highly unlikely for a meltdown to occur.

Most cores that use this fuel are "high leakage" cores where the excess leaked

neutrons can be utilized for research. TRIGA fuel was originally designed to use highly

enriched uranium, however in 1978 the U.S. Department of Energy launched its

Reduced Enrichment for Research Test Reactors program, which promoted reactor

conversion to low-enriched uranium fuel. A total of 35 TRIGA reactors have been

installed at locations across the USA. A further 35 reactors have been installed in

other countries.

Actinide Fuel:

In a fast neutron reactor, the minor actinides produced by neutron capture of

uranium and plutonium can be used as fuel. Metal actinide fuel is typically an alloy of

zirconium, uranium, plutonium and the minor actinides. It can be made inherently safe

as thermal expansion of the metal alloy will increase neutron leakage.

4.2.3 Ceramic Fuels:

Ceramic fuels other than oxides have the advantage of high heat conductivities and

melting points, but they are more prone to swelling than oxide fuels and are not

understood as well.

Uranium Nitride:

This is often the fuel of choice for reactor designs that NASA produces, one

advantage is that UN has a better thermal conductivity than UO2. Uranium nitride has

a very high melting point. This fuel has the disadvantage that unless 15N was used (in

place of the more common 14N) that a large amount of 14C would be generated from

the nitrogen by the (n,p) reaction. As the nitrogen required for such a fuel would be so

expensive it is likely that the fuel would have to be reprocessed by a pyro method to

enable to the 15N to be recovered. It is likely that if the fuel was processed and

dissolved in nitric acid that the nitrogen enriched with 15N would be diluted with the

common 14N.

Uranium Carbide:

Much of what is known about uranium carbide is in the form of pin-type fuel

elements for liquid metal fast breeder reactors during their intense study during the

'60s and '70s. However, recently there has been a revived interest in uranium carbide

in the form of plate fuel and most notably, micro fuel particles.

16

The high thermal conductivity and high melting point makes uranium carbide an

attractive fuel. In addition, because of the absence of oxygen in this fuel (during the

course of irradiation, excess gas pressure can build from the formation of O2 or other

gases) as well as the ability to complement a ceramic coating (a ceramic-ceramic

interface has structural and chemical advantages), uranium carbide could be the ideal

fuel candidate for certain Generation IV reactors such as the gas-cooled fast reactor.

4.2.4 Common physical forms of nuclear fuel:

PWR Fuel:

Pressurized water reactor (PWR) fuel consists of cylindrical rods put into

bundles. A uranium oxide ceramic is formed into pellets and inserted

into Zircaloy tubes that are bundled together. The Zircaloy tubes are about 1 cm in

diameter, and the fuel cladding gap is filled with helium gas to improve the conduction

of heat from the fuel to the cladding. There are about 179-264 fuel rods per fuel bundle

and about 121 to 193 fuel bundles are loaded into a reactor core. Generally, the fuel

bundles consist of fuel rods bundled 14×14 to 17×17. PWR fuel bundles are about 4

meters long. In PWR fuel bundles, control rods are inserted through the top directly

into the fuel bundle. The fuel bundles usually are enriched several percent in 235U. The

uranium oxide is dried before inserting into the tubes to try to eliminate moisture in the

ceramic fuel that can lead to corrosion and hydrogen embrittlement. The Zircaloy

tubes are pressurized with helium to try to minimize pellet-cladding interaction which

can lead to fuel rod failure over long periods.

BWR Fuel:

In boiling water reactors (BWR), the fuel is similar to PWR fuel except that the

bundles are "canned"; that is, there is a thin tube surrounding each bundle. This is

primarily done to prevent local density from affecting neutronics and thermal hydraulics

of the reactor core. In modern BWR fuel bundles, there are either 91, 92, or 96 fuel

rods per assembly depending on the manufacturer. A range between 368 assemblies

for the smallest and 800 assemblies for the largest U.S. BWR forms the reactor core.

Each BWR fuel rod is back filled with helium to a pressure of about three atmospheres

(300 kPa).

17

5. VARIOUS PLANTS AT BARC TARAPUR

5.1 PREFRE (POWER REACTOR FUEL REPROCESSING) PLANT:

The PREFRE

plant located at Tarapur

and commissioned in 1975

reprocesses zircaloy clad

oxide spent fuel using

chop-leach technique for

the head end.

Besides providing

the required Plutonium,

several campaigns of

reprocessing have also

been carried out under

international safeguards in

this plant, thereby,

providing valuable experience in material accounting practices adhering to the

international standards.

Around 40 years of experience in the spent fuel reprocessing based on PUREX

process has given the confidence that this technology can be successfully employed for

the recovery of both U and Pu with yield exceeding 99.5%. Substantial reduction in waste

volume has been achieved over the years by resorting to salt free reagents. Evaporation

followed by acid reduction by formaldehyde is used to reduce the high level waste

volume. The overall decontamination factors for the Pu and U products from fission

products exceed 106 and are handled subsequently with minimum radiation protection.

5.2 SSSF (SOLID WASTE STORAGE SURVEILLANCE FACILITY) PLANT:

BARC had

established a waste

treatment plants at

Tarapur site also had

waste treatment plants

to take care of the

wastes arising from

operations of TAPS and

PREFRE. All these

plants came under the

18

review of SARCOP from 1987 onwards.

Management of HL waste has to take into account the need for their isolation

and surveillance for extended periods of time. To meet this objective in the long term

perspective, waste isolation system comprising multiple barriers are employed. The

vitreous matrixes in which these waste are immobilized constitute the primary barrier.

This along with its packaging, engineered barriers in the repository and the surrounding

geology (secondary barriers) together are expected to prevent the recycling of

radionuclides back into human environment so as to pose no hazard. The long term

strategy for management of High Level Waste would involve partitioning of long lived

radionuclides that would result in reduction of radioactivity. Ceramic matrices are also

being pursued to address specific waste streams.

5.3 AFFF (ADVANCED FUEL FABRICATION FACILITY):

Fig. 5.3: AFFF Plant at BARC

BARC installed an Advanced Fuel Fabrication Facility (A3F) at Tarapur in 1989

for fabrication of mixed oxide (MOX) fuel subassemblies for Tarapur Atomic Power

Station (TAPS). AERB constituted an ACPSR with K. Balaramamoorthy, the then Chief

Executive, and NFC to carry out the safety review of the project. ACPSR had several

detailed discussions on the engineered safety features of the plant to ensure the

containment of radioactivity during plutonium powder handling operations, criticality

safety etc. Based on the recommendations of this Committee, AERB issued the

authorization

For regular operation of A3F in 1994.

19

6. MECHANICAL WORKSHOP AT AFFF, BARC

6.1 INTRODUCTION:

This Work Shop is engaged with fabrication of various components and

maintenance in the fuel fabrication process and tries to keep all the machines in working

conditions, with higher availability factor, the various operations are carried such as lathe

operation i.e. turning, milling, shaping and all the utility operations.

6.2 WORKSHOP:

As a mechanical engineer, it is required to know about different type of the

machines tools i.e. Lathe, Milling, Shaping machine etc., and their specifications &

applications. The entire above requirement is completely fulfilled by workshop. Workshop

consists of almost all types of machines workshop gives the practical information about

the various operations carried on different machines following are some important

machines of the workshop.

Lathe machine

Milling machine

Shaper machine

Rolling machine

Power press (Mechanical type)

Shearing machine

All cut machine ( Band Saw )

Grinding machine (cylindrical, flexible shaft grinders, surface grinder)

Tool cutter grinder machine

Radial drilling machine

Welding machine (ARC, TIG, PLASAMA)

6.2.1 LATHE MACHINE:

LATHE MACHINE is the most important machine tool. The main function of

the lathe machine is removing the material to provide desired shape on the job. This

can be accomplished by revolving the job against the single point cutting tool. There

are different types of lathe machines.

Speed Lathe

Engine Lathe

Capstan and Turret Lathe

20

Fig.6.2.1: Lathe Machine

DIFFERENT TYPE OF LATHE OPERATIONS AS FOLLOWS:

Plain Turning

Step Turning

Taper Turning

Threading

Grooving

Drilling

Reaming

SPECIFICATION:

Bed length – 2500 mm

Nominal height – 220 mm

Holding length – 1800 mm

Feed – 0.04-2.25 per rev

Speed – 40-2040 rpm

21

6.2.2 MILLING MACHINE:

Fig.6.2.2: Milling Machine

Milling is the process of removing the metal by feeding the work past at

rotating multipoint cutter.

Classification:

Column & knee type

Bed type (simplex, duplex, triplex)

Plano type

Special purpose machine

In the workshop there is a horizontal milling machine of column & knee type

with a vertical attachment. Principal parts of column & knee type machine are as

follows.

Motor, over arm, cutter, spindle, arbor, column table, saddle knee, elevating

screw base etc.

22

Specifications:

Table size - 1350 x 350 mm

Table traverse - both manually & automatic

Longitudinal - 810mm & 800mm

Cross - 220mm & 235mm Vertical - 340mm& 360mm Main motor - 3.7kw (5Hp)

6.2.3 SHAPING MACHINE:

Fig.6.2.3: Shaping Machine

The shaper is a reciprocating type of machine tool intended primarily to

produce flat surface. These surfaces may be horizontal, vertical or inclined. In

general, the shaper can produce any surface composed of straight line element.

In modern machine shop there is a universal shaper, in this shaper swiveled

about an axis parallel to the ram ways and the upper portion of the table can be tilted

about a second horizontal axis perpendicular to the first axis as the work mounted on

the table can be adjust in different planes, the machine is most suitable for different

types of work like producing flat surfaces, internal or external key cutting etc.,

23

6.2.4 Centreless Grinding machine:

Fig.6.2.4: Centreless Grinding Machine

Centreless grinding is an OD (outer diameter) grinding process. In difference

from other cylindrical processes, where the work piece is held in the grinding machine,

while grinding between centers, the work piece is not mechanically constrained during

centreless (centerless) grinding. Therefore the parts to be ground on a centreless

(centerless) grinder do not need center holes, drivers or work head fixtures at the

ends. Instead, the work piece is supported in the grinding machine on its own outer

diameter by a work blade and by the regulating wheel. The work piece is rotating

between a high speed grinding wheel and a slower speed regulating wheel with a

smaller diameter.

G: Grinding Wheel - R: Regulating Wheel - B: Blade - W: Work piece

24

The blade of the grinding machine is usually positioned in a way that the

center of the work piece is higher than the virtual line between the centers of the

regulating wheel and the grinding wheel. Also the blade is designed with an angle in

order to ensure that the work piece is fixed between the blade and the regulating

wheel. The regulating wheel consists of soft material like rubber and can contain some

hard grain material to achieve good traction between work piece and regulating wheel.

ROUNDNESS:

Centreless (centerless) grinding can perform excellent roundness of the work

piece. However, caused by the simultaneous suspending and machining of the work

piece surface it is possible that process typical roundness errors are generated.

Proper adjustment of the grinding machine and the grinding slot geometry is essential.

When a high spot comes in contact with the regulating wheel, then on the other side of

the work piece a low point will be ground. However this low point must not be exactly

in the opposite side of the work piece. The grinding machine has to be set up in a way

that a polygon form is ground with so many corners that it is almost round finally.

APPLICATIONS:

Mass Production:

E.g. bolts, shafts, bearings, hubs, valves, needles, axles, pivots

6.2.5 GRINDING MACHINE:

There are various types of grinding machines available in the workshop. As

required to their applications. Such as hydraulic cylindrical grinding machine precision,

hydraulic surface grinding machine, and portable grinding machine.

Selection of grinding wheel:

Type of

Bond

Grinding

Operation Abrasive Size Grade

Vitrified precision

grinding

aluminum

oxide 36 to 80 N to J

Vitrified precision

grinding

Silicon

Carbide 36 to 80 I to L

Aluminum oxide is use for grinding of steels, hard bronze, manganese, nickel,

chrome, and phosphor bronze.

Silicon carbide is applicable for the grinding of aluminum, soft brass, and

bronze, cast iron, copper, nickel, chrome, tungsten carbide and rubber.

25

Specification:

Ac / Dc - 230 volts

N/L rpm - 11500 and 1200

F/L amps - 1.74 A

Duty cycle - 30 min

SSppeecciiffiiccaattiioonn ooff GGrriinnddiinngg wwhheeeell:: ““WW AAAA 6600 KK 55 VV 1100””

WW :: -- PPrreeffiixx ((MMaannuuffaaccttuurree‘‘ss CCooddee))

AAAA:: -- AAlluummiinnuumm ooxxiiddee

6600::-- GGrraaiinn SSiizzee MMeeddiiuumm

KK :: -- GGrraaddee MMeeddiiuumm

55 ::-- SSttrruuccttuurree DDeennssee

VV:: -- BBoonndd VViittrriiffiieedd

1100::-- SSuuffffiixx

FFiigg..66..22..55 GGrriinnddiinngg WWhheeeell

## SSttrruuccttuurree ooff GGrriinnddiinngg WWhheeeell::

DDEENNSSEE 11 22 33 44 55 66 77 88

OOPPEENN 99 1100 1111 1122 1133 1144 1155

## GGrraaddee ooff GGrriinnddiinngg WWhheeeell

SSoofftt AA BB CC DD EE FF GG HH

MMeeddiiuumm II JJ KK LL MM NN OO PP

HHaarrdd QQ RR SS TT UU VV WW XX YY ZZ

## GGrriitt GGrraaddee SSttrruuccttuurree ooff WWhheeeellss

CCooaarrssee 1100 1122 1144 1166 2200 2244

MMeeddiiuumm 3300 3366 4466 5544 6600

FFiinnee 8800 110000 112200 115500 118800

VVeerryy

ffiinnee

222200 224400 228800 332200 440000 550000 660000

26

6.2.6 ROLLING MACHINE:

The rolling machine is one of the important machines. It works only on

applying pressure. In workshop there are two rolling machines available out of which

one is manually operated. The entire structures of both the machines are made up of

steels. Frame is fully welded structure made of heavy section steels.

Working principle:

The desired length of metal is past between the two rollers and simultaneously

rotating rollers by manually or through power which applies pressure on metal plate to

get circular shape.

6.2.7 POWER SHEARING MACHINE:

Fig.6.2.7: Power Shearing Machine

27

Working:

A punch (or moving blade) is used to push the work piece against the die (or

fixed blade), which is fixed. Usually the clearance between the two is 5 to 10% of the

thickness of the material, but dependent on the material. Clearance is defined as the

separation between the blades, measured at the point where the cutting action takes

place and perpendicular to the direction of blade movement. It affects the finish of the

cut and the machine's power consumption. This causes the material to experience

highly localized shear stresses between the punch and die. The material will then fail

when the punch has moved 15 to 60% the thickness of the material, because the

shear stresses are greater than the shear strength of the material and the remainder of

the material is torn. Two distinct sections can be seen on a sheared work piece, the

first part being plastic deformation and the second being fractured. Because of normal

inhomogeneities in materials and inconsistencies in clearance between the punch and

die, the shearing action does not occur in a uniform manner. The fracture will begin at

the weakest point and progress to the next weakest point until the entire work piece

has been sheared; this is what causes the rough edge. The rough edge can be

reduced if the work piece is clamped from the top with a die cushion. Above a certain

pressure the fracture zone can be completely eliminated.[3] However, the sheared

edge of the work piece will usually experience work hardening and cracking. If the

work piece has too much clearance, then it may experience roll-over or heavy burring.

Specification:

Model Cutting

Thickness

Cutting

Length

Stroke

/ Min

Table

Size (W

x L)

Blade

Size (W

x T x L)

No. of

Blades

Motor

Power

Height of

Table

from

Floor

UCS Mm Mm nos mm mm nos HP /

HPM mm

04075 4 750 65 300 x

1380

18 x 55 x

770 2

5 /

1440 815

6.2.8 COOLANT:

Coolant is a solid or liquid flowing medium which is used for the cooling of job

& tool for removing the heat which is produced between job & tool due to its friction

while performing machining like turning, milling, drilling etc.

Purposes:

To increase the tool life

To improve the surface finishing

28

To flow away the chips which is formed during the operation on the job?

Classification:

Solid coolant e.g.- Graphite

Liquid coolant e.g. - Water, Soluble oil etc.

Semi Solid coolant e.g.- Wax

Here in workshop we are using soluble oil cool cut 40 with the ratio of 1:20

with water. This coolant is used for stainless steel, mild steel etc. For aluminum &

copper kerosene is used as a coolant. For cast iron & brass there is no requirement of

coolant.

6.2.9 TOOL: Tool is a substance which is used for removing the excess metal from the

work piece in the form of chips to get required shape and size with the help of its

cutting edge. The tool material should be harder than the work piece on which

operation has to done.

In the workshop we are using High Speed Steel Tool & Carbide Tool

COMPOSITION:

High Speed Steel:

Carbon - 0.70-1.50 % Chromium – 4.00-4.50 %

Vanadium – 1.00-5.00% Tungsten - 12-20%

Fig.6.2.9 (a): High Speed Tool

Carbide:

Tungsten base

Carbon – 40% Tungsten - 60%

Cobalt base

Cabon-40% Tungsten & Cobalt – 60%

29

7. WWEELLDDIINNGG

Welding and cutting operation are frequently used in engineering industry in

fabrication, repair and maintenance work. Welding is a process to unite pieces of metal at

a joint faces by heat and use of a filler material. Cutting is a process to remove the metal

by chemical reaction of the metal at high temperature. In both these operation, the

common factor is high heat energy and high temperature for melting or fusing of metals.

Types of welding used in workshop are

SMAW

TIG welding (GTAW)

7.1 TIG WELDING (GTAW)

Fig.7.1: GTAW Welding

It is an arc welding process in which the heat is produced between a non-

consumable electrode and the work metal. A stream of a gas or a mixture of gases

protects the electrode weld pool arc and adjacent heat areas of the work piece. The gas

shield must provide full protection even small amount of entrained air can contaminate

the weld.

Because of the non-consumable electrode, a weld can be made by fusion of

the base metal without addition of filler metal. A filler may be used, however depending

on the requirement that have been established for the particular joint TIG welding is an all

30

Position welding process and is especially well adapted to the welding of thin metals

often as thin as 0.005 inch.

7.1.1 Metals that can be welded: The nature of GTAW permits its use for welding of most metals and alloys.

Metals that are gas tungsten arc welded includes carbon and alloy steels, stainless

steels, heat resistant alloys, refractory metals, aluminum alloys, copper alloys,

magnesium alloys, nickel alloys etc.

7.1.2 Technical chart for TIG welding

Thickness

(mm) Electrode Filler rod

Current

(amps)

Gas flow

(psi)

1.5 1.6-2.4 1.6 88-130 4-6

3 2.4-3.2 1.6 120-224 4-6

6 3.2-4.8 3.2 220-350 6-8

12 4.8 3.2-4.8 330-420 6-10

7.1.3 Technical chart for cutting process for Argon gas

Thickness

(mm)

Current (amps) Gas flow (psi)

5-8 100-150 10-15

8-16 150-200 15-20

16-25 200-250 20-25

25-32 250-300 25-35

Argon is one of the excellent gases that freely give up electrons, hence

produce a more stable and quite arc during welding. Stable arc reduces the spatter

effect. But argon gas cannot be used for deeper penetration. This gas prevents direct

contact of air with weld pool

31

7.2 SMAW (SHIELDING METAL ARC WELDING:

Fig. 7.2: SMAW Welding

It is a manual arc welding process in which the heat for welding is generated

by an arc established between a flux coated consumable electrode and a work piece.

The electrode tip, molten weld pool and adjacent areas of the work pieces are protected

from atmosphere. Contamination by gaseous shield obtained from the consumption and

decomposition of the electrode covering. Additional shielding is provided for the molten

metal in the weld pool, by a covering of the molten metal in the weld pool by a covering of

molten flux or slug.

7.2.1 Join quality and strength:

The quality and strength of the shielded metal arc welded joints can be

controlled as easily as the quality and strength joints welded by other manual methods

that use consumable electrodes. SMAW electrode material is available to the match

the Properties of most ferrous base metals, allowing the properties of a joint to match

those of alloys jointed.

7.2.2 Metals commonly welded:

By SMAW arc carbon and low alloy steels, stainless steel, heat resistant

alloys, cast iron and high strength and harden able steels can also be shielded metal

arc welded, but process that includes preheating, post heating or both may be needed.

32

7.3 WELDING ELECTRODE SPECIFICATION:

Electrode is a metallic rod coated or non-coated, consumable or non-

consumable used to transmit heat that melts the work piece or base metal by generating

arc. The type of electrode required is depending upon the following factors.

The strength required

Type of base metal to be welded

Additional filler metal is used or not.

Size of electrode required will depend upon the following factors.

Type of joint and gap to be bridge between two plates to be welded

Amount of current supplied.

Classification of electrode:

Welding electrode

Selection of electrode

Chemical composition

Thickness of the work piece

Nature of electrode coating positions

Non consumable

Carbon graphite Tungsten

Bare electrode

Flux cored electrode

Pure electrode

Consumable

Thoriated electrode Zicroniated electrode

33

7.4 WELDING DEFECTS:

7.4.1 Cracks:

Causes

Poor ductility of base metal

Fast arc travel speed

Internal stresses in the base metal

Remedies

Metal should be tested for its composition as well as mechanical properties

before use and the welding speed must be up to desired value.

7.4.2 Distortion:

Causes

Very high welding speed

Continuous welding

Residual stresses in the base metal

Remedies

Use of jigs, fixture, clamps, may minimize the distortion.

Distortion can also minimize by proper weld tacking.

7.4.3 Incomplete penetration:

Causes

Fewer arcs current

Faster arc travel speed

Two large electrode diameters

Wrongly held electrode

Remedies

In case of filling material to the corner box, sufficient temperature is required to

melt the base metal as well as electrode. Hence arc current and voltage should be

proper. Position of the electrode should be near about an angle of 70 to 80 degree the vertical

plane, Electrode of proper size and shape should be used.

34

7.4.4 Slag conclusion:

Causes

Too high and too low arc current.

In sufficient chipping and cleaning off previous passes in multi pass

welding.

Improper selection of electrode coated with flux.

Remedies

Set the current and voltage at desired value

Use of proper flux coated electrode

Cleaning of slag after welding

35

8. ENCLOSURE BOX PANEL

8.1 INTRODUCTION:

An enclosure box is cubical / cuboidal shape box or container which confines

the radioactive materials or toxic chemical which contaminates the surroundings

environment. There are mostly six types of standard enclosure and there used as per the

required processes & geometry of the process equipment.

1) Type I (1094 x 637 x 1094) mm

2) Type II (1731 x 637 x 1094) mm

3) Type III (2188 x 637 x 1094) mm

4) Type IV (1094 x 1094 x 1094) mm

5) Type V (1731 x 1094 x 1094) mm

6) Type VI (2188 x 1094 x 1094) mm

8.2 PANEL: Panel act as protective shield which is attached to enclosure box from all side

& makes the enclosure leak tight to required level. It protect the surroundings from the

contamination .It has number of parts as per use purposes.

There are four types of panel in an enclosure box which are as follows:-

Front Panel

Back Panel

Top Panel

Side Panel

8.2.1 FRONT PANEL:

In a front panel four glove ports and one viewing window. The viewing window

will be rectangular and circular. It is as per the viewing aspect of working condition.

8.2.2 BACK PANEL:

Back panel is made up of stainless steel metal or glass& mostly only viewing

window is provided & if required as per condition ports can be provided if required.

36

8.2.3 TOP PANEL:

The top panel is also made of stainless steel on which a large size of

rectangular glass panel is provided by which light can pass for good visibility, outlet

port is provided for exhaust in enclosure box & other required accessories are

provided if required on the top panel for enclosure box.

8.2.4 SIDE PANEL:

In a side panel a transfer port is provided by which radioactive material is

transfer for one enclosure box to other enclosure box for other treatment or processes.

The panels are designed ergonomically to suit the day to day operations &

maintenances requirements. The positions of parts & windows are decided on the basis

of comfort of operations & maintenance, ease of accessibilities & visibility to maximum

possible extents.

The flatness & the geometry of these panels are very critical to get the leak

tightness point of view.

8.3 DESIGN CONCIDERATION:

For fabrication of panel stainless steel of grade 304 has to use because it has high

corrosion resistance & its density is sufficient to minimize the penetration rate of

radioactive radiation by less thickness of sheet only.

The stainless steel of 3mm thick sheet is used for fabrication of panels because in

the system maximum up to negative 4‖ of WG pressure with respect to working area

can be created at critical condition so at that pressure it should not rupture.

In panel four port holes should be provided to cover max. Area in the enclosure box

for any operation or maintenance. The diameter of port hole & distance between two

ports is should be according to the normal human height & shoulder width for easy in

work.

The viewing window should be design to cover maximum visibility area in the

enclosure box. Generally circular window is used but if it is not sufficient than

rectangular window has to provide.

8.4 SELECTION OF MATERIAL:

8.4.1 STAINLESS STEEL: It is alloy of iron with a minimum of 10.5% chromium, chromium produces thin

layer of oxide on the surface of the steel know as passive layer. This prevents any

further corrosion of surface, increasing of chromium gives an increased resistance of

corrosion.

37

Type Of Stainless Steel

Ferritic stainless steel

Austenitic stainless steel

Martenstic stainless steel

Duplex stainless steel

Precipitation Hardening stainless steel

Ferritic Stainless Steel

These stainless steel have better engineering properties than austenitic grade

but have reduced corrosion resistance because of the lower chromium & nickel

content.

Martenstic Stainless Steel

It is extremely strong & tough but not as corrosion resistant then the other

class of stainless steel.

Duplex

It is the combination of 50% austenite steel & 50% ferrite Steel.

Precipitation Hardening

These stainless steel can develop very high strength by adding element such

as copper, niobium & aluminum to the steel corrosion resistance is comparable to

standard austenite.

Austenitic Stainless Steel

This stainless steel are shown 300 series, Stainless Steel have an Austenitic

crystalline structure which is an FCC face centered cubic crystal structure, it make up

over 70% of total SS production They contain max 0.15% carbon, min 16% chromium

&sufficient nickel & manganese to retain as austenite structure at all temp. From the

cryogenic region to m.p. of the alloys.

In this industries we are dealing with the toxic material not very high stresses

or load so in this field we can use low strength material but we required high corrosion

resistive material & less cost material so for that austenitic stainless steel is good to

use.

Therefore glove box & panels are mostly made of stainless steel of grade 304.

This stainless steel possess in austenitic type of Stainless Steel.

Composition of SS304

Carbon – 0.08% Magnesium – 2.00%

Silicon – 1.00% Chromium – 18.00-20.00 %

Nickel – 8.00-10.5 %

38

Properties Of Stainless Steel

High density because of which radiation cannot penetrate.

It is a non-corrosive material.

Surface finishing is very high & it property of hardness is high due to which

scratch can be avoided.

Ease of cleaning due to better finish.

8.5 FABRICATION OF PANEL: [[

First with the help of power shearing machine the sheet of 3mm thick is cut of

dimension 945 x 850 mm & the corner radius of 103 mm is also cut by power shearing &

after that the corner radius is ground to get required finishing & dimension. For this refer

plan of panel on drawing.

Now in panel there are two port holes are provided of diameter 204 mm which

is done by trepanning operation by using trepanning tool on the drill machine. For this

refer plan of panel & section D-D on drawing.

Now one viewing window is provided of dimension 750 x 150 mm which is cut

by bend saw machine & the corner radius of 30 mm is done by this machine only. For

this refer plan of panel & section X-X on drawing.

Now the MS bright bar is fabricated by bending the bar to required angle this

is done by a fixture which is made of required angle for this refer section X-X on drawing.

This bar is of material mild steel because on MS bending, tapping operation

are easy to be done as compare to SS & it is outside of enclosure box so there is no

contact between radioactive material so no chance of contamination & it also reduce cost

of production.

Now MS bright bar of dimension 15 x 13 mm is welded on the viewing window

by GTAW welding process for screwing the glass. This welding become dissimilar

welding so for that the filler metal is used is of grade SS309.

Now glove ports are made from 8‖ NB schedule 40 pipe of material SS304

which means pipe outer diameter is 219.1mm & wall thickness is 8.18mm & then we

fabricate three ‗O‘ ring groove on the pipe with the help of radius grooving operation on

very low rpm by using radius tool on the lathe machine, for this refer section A-A on

drawing.

This O ring is provided on the surface of glove port to get the proper sealing

between glove port & gauntlet & its surface finishing is 0.025 to 1.6 micron.

Now these ports are welded on port hole with the help of GTAW welding

process. This welding is a similar welding so in this the filler wire used is of grade SS308.

The whole GTAW welding process used in the plant we are using argon gas for shielding

& 2% thoriated Tungsten non consumable electrode which is also called red tip electrode

according to AWS.

39

By clamping strip the glass on viewing window & panel on enclosure box is

clamped by screwing. That clamping strip is fabricated by die & punch, for this refer

section X-X on drawing.

On this clamping strip the hole is drilled of required dimension for screwing the

socket head M6 x 15. This whole diameter is given by-

[D – 1.3 p]

Where D = Diameter of screw

P = Pitch which is taken mostly 1 mm

For leak tightness we are using Neoprene gasket of width 12 x thick 1.6 mm &

Neoprene O ring of dia. 6.99 mm which is placed between the glass panel & clamping

strip & it also placed in between panel & enclosure. In gasket there is a surface contact

between the two surfaces & In O ring there is a line contact between the two surfaces. So

O ring gives better leak tightness.

8.5.1 Welding inspection method:

Here after any welding like port with port hole, MS bright bar with viewing

window etc. in the panel the welding inspection has to be done to find out any defect

which are not visible in visual inspection which are done by naked eyes.

Types Of Welding Inspection Test

Liquid Penetrant Test

Magnetic Particle Test

Radio graphical Test

Ultrasonic Test etc.

But here we are using only liquid penetrant test. This test can only indicate surface

defect, if defect is in the inside the surface then it cannot deduct it. But here we are

more concern about leak proof more than strength wise so this test is sufficient.

8.5.2 Principle of the Die Penetrant test:

D.P test is working on the principle of capillary action. Following is the

procedure to carry out the D.P test.

First of all, cleaning of welding joint is required to be done. This will remove all

the dust particle, grease, oil etc. cleaning process is done with ACETONE.

Once the cleaning is completed, dye is sprayed on the welding joint whose

defects are to be checked. Dye is also known as penetrating. It is red in color. As dye

penetrates in the cracks on the welding joint or surface of metallic enclosure, it is kept

for 15 min to 20 min.

After this, it is again clean with wet cloth. This will help to remove penetrant

from the surface of the component. This process is carried out till there is no penetrant

40

on the surface. Here penetrant which is penetrating inside the cracks of welding joint

are not cleaned by acetone.

Then white developer is sprayed on the surface. This developer gives the

capillary action, and due to that action the pink penetrant that is inside the cracks is

shown on the surface in the form of line, or spot.

Once we found these pink lines or cracks on the surface we may conclude that the

cracks are present on the welded material. And try to remove it.

8.5.3 LEAK TESTING OF PANELS:

After complete dimensional inspection & assembly of panel, enclosure box

etc. Then we go for pressure drop testing method for leak testing of panel & enclosure

box. Where we pressurize the enclosure box up to required testing standard & then by

calculation we find the leak rate.

Calculation

P1 & T1 = Initial pressure & temp.

P2 & T2 = Final pressure & temp.

V = volume is constant

We know,

PV= nRT

Where, R = Universal gas constant

n = no. of mole

Initial condition P1V = n1RT1

Final condition P2V = n2RT2

Leak Rate = n1 – n2/n2 x 100/no. of hrs

OR

Leak Rate = (P1T2 – P2T1)/P1T2 x 100 / no of hrs

41

CONCLUSION In pursuit of the peaceful uses of Atomic Energy, power generation based on nuclear

energy assumes first and foremost place and India has achieved many milestones in this

area. A well planned programme for the progressive expansion for the tapping of atomic

energy for electricity keeping in view of the country‘s future requirements for increased

power generation capacity and available resources has been under implementation. A

strong R&D base has been established and functions as a back bone for the smooth

transition of the research and development activities to the deployment phase and thereby

realising the Department of Atomic Energy‘s mandate. Many technologies of strategic

importance have been mastered to meet developmental needs. Indigenous technology

development in the areas of fuel reprocessing, enrichment, production of special materials,

computers, lasers, accelerators represents a whole spectrum of activities necessary for

realising full potential of our energy resources to meet future energy needs. Radiation

Technology and Isotope Applications represents another prominent area of the peaceful

uses of Atomic Energy in health care, agriculture, industries, hydrology and food

preservation where self- reliance has been accomplished.