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
2
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
3
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
4
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.
6
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:-
7
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.
9
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.
11
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.
Recommended