India's progress in fusion energy.

[Trololo]

Idk why but our heavy engineering products look so 1970s Soviet Union-ish.

 

[Karthi]

42 GHz Gyrotron System

ADITYA Dismantled at 2015.

very old pics of ADITYA TOKOMAK

ADITYA Poloidal Limiters inside the vacuum Vessel

ADITYA Vacuum vessel at L&T facility Mumbai an old pic

ADITYA Upgraded

Upgraded vaccum vessel Godrej

 

[Karthi]

ADITYA U Vaccum Vessel

ECRH System in ADITYA

LH Grill Anteena

Toroidal Field Coils

Toroidal Belt Limiter

Vacuum Vessel ADITYA U

Vacuum vessel Support Structures

 

[Karthi]

Startup plans nuclear fusion plant in India.

Akash Singh, an Indian-origin entrepreneur based out of Silicon Valley, is all set to venture into nuclear fusion by early next year. Singh is working on 'Project Sanlayan', which is supported by a team of international investors. This was revealed at the Bihar Science Conference held early this month.

The project, which will be a first-of-its-kind initiative by any public or private entity in India, aims to commercialise nuclear fusion for power production by 2035. The location for setting up the project in India, however, is yet to be finalised.

"Our objective is to start a nuclear fusion R&D plant in India in early 2021. We want to use the clean energy which will be highly useful for peaceful purposes in India and around the world," Singh said,

It assumes significance as the people around the world are still working on such projects and they are likely to become operational in a minimum 15 years from now. For example, the UK government plans to set up a nuclear fusion plant by 2040 and has asked societies/communities to provide land for the same by March.

The Indian government has formed a separate entity to support the private industry in space, named as The Indian National Space Space Promotion and Authorisation Centres (IN-SPACE) to promote and guide private industries in space activities through encouraging policies and a friendly regulatory environment. The question arises if they will do it for the nuclear fusion sector, too?

Interestingly, India can develop technology at 1/15th to 1/20th cost, compared to the western world like it did in the IT sector as well as the space sector in the past.

Supported by the Indian government, Anant Technologies has become the first Indian firm to tap the global space markets and it is working on the project jointly with Saturn Satellite.

"It is a momentous occasion for a private initiative in the field of nuclear fusion in India," says Dr Prabhat Ranjan, a nuclear fusion scientist and the former executive director of TIFAC. Ranjan is currently working as the Vice Chancellor at D Y Patil International University in Pune. He is also a mentor of Project Sanlayan.

The document 'Technology Vision 2035', which was unveiled by Prime Minister Narendra Modi in New Delhi on 3 January, 2016, has a mention of the nuclear fusion. Apart from clean and long-term energy, the nuclear fusion project is also likely to contribute to the field of medical isotope and space propulsion.

 

[Karthi]

ADITYA-U resumed operations from 24 November, 2020 and continued till 18 th December, 2020. Almost 227 discharges were obtained with varied toroidal magnetic field (B) from 0.9 Tesla to 1.07 Tesla. The extensive wall conditioning with Hydrogen GDC as well as combination of Argon + Hydrogen GDC showed substantial reduction in various impurities mass species like H2, 02 N2 etc.
The longest ever ADITYA-U discharge of ~ 386 ms with ~ 134 kA of peak plasma current (Shot #34009) has been achieved by using 70% of available Volt-Sec at ~ 19 V of peak loop voltage. The discharge was obtained at toroidal field of ~ 1.07 Tesla and in absence of any strong pre-ionization.

Aditya-Upgrade plasma discharge with divertor coil operation showing strike points

actual experimental photograph, showing ion recirculation across the cathode openings.

Cylindrical IEC fusion device

The spherical ICE fusion device


Kinetic simulations have been performed using particle-in-cell method to analyse the behaviour of ions inside a cylindrical inertial electrostatic confinement fusion (IECF) device which is being developed as a table top neutron source. The ion recirculation across the gridded cathode, ion density and potential well profiles are investigated in a discharge plasma using XOOPIC (X11- based object-oriented particle-in-cell) code. The simulated results are also compared with the experiment to benchmark the results. The recirculation of the ions can be visualised from the phase space during runtime.The simulation contour plot of ions during -1 kV cathode voltage clearly indicates the signature of recirculating ions across the cathode grid openings. This compares well with the photograph of the actual experiment. Ion spokes or channels can be observed to
be coming out from the central core region. Ion density profile was measured along with experimental profile using Langmuir probe, and a maximum ion density is found to be ~10 16 m-3 during -5 kV operation. The potential profile indicates the formation of multiple or double well structures during -5 kV cathode voltage operation

The BEAM experimental device


plasmas confined by toroidal magnetic field is a viable mechanism to achieve controlled thermonuclear
fusion in laboratories from which abundant electricity can be produced. As the ITER fusion device is reaching towards achieving its first plasma, the nation is concurrently building technologies that will be required to indigenize a fusion reactor.

From both physics and engineering perspective, the laboratory experiments are highly needed to perfect the underlying technologies, create the absolute knowhow and generate relevant experimental data for benchmarking the modelling results associated with these complex machines.

In the Magnetized Plasma Development Laboratory, a Basic Experimental setup with Axial Magnetic field (BEAM) has been developed in which rigorous experiments are carried out on positive and negative ion beams produced by diverse discharge configurations; with primary focus on their interactions with electrodes in magnetized plasma. The BEAM setup is equipped with a quadrupole mass and energy analyzer system for diagnostics. Using this setup, energy distribution of plasma ions, multi charge ionic species, energetic neutrals that may be generated using different types of plasma sources such as hall thruster, high voltage sputtering magnetron discharge, ion extraction using plasma grids etc. can be characterized. The device is also equipped with a suite of indigenously developed electric probe diagnostic system that provide accurate measurement of plasma parameters at various axial and radial location inside the BEAM setup.

Electrode assembly of Capacitive-Coupled RF discharge

Inside view of BEAM setup showing the cylindrical discharge electrodes

Ring-shape hot plasma produced by cylindrical RF discharge with axial magnetic field.


Recent experiments in the BEAM demonstrates a remarkable mechanism of controlling the radial electron temperature and plasma uniformity inside cylindrical plasma column by a combination of external
plate biasing and diverging magnetic field.

This effect is highly promising for applications in industries, where homogeneous plasma is required for the treatment of large area substrates. The uniform plasma is also required in plasma based ion sources to minimize the beam divergence and spatial energy distribution of ions extracted from the source. Controlling the electron temperature is also important for the production of negative ions. This has been achieved in the BEAM by a suitable combination of electrode geometry and axial magnetic field. Due to high neutralization efficiency, negative ions are used for the neutral beam generation for applications ranging from plasma thruster to neutral beam heating in fusion device, and also for weapons. Presently studies on charge particle transport across magnetic field lines, wake creation in a flowing/ non - flowing magnetized plasma, investigation of radio-frequency magnetized sheaths and plasma surface interactions are some of the work being carried out using this device. In the future, it is planned to install a new set of electro -magnets in the BEAM to achieve peak magnetic field of 300 Gauss. The facility will lay
emphasis on development indigenous ion beam sources.

 

[Karthi]

After 25 years of successful operation of ADITYA tokamak (~ 30,000 discharges) in a circular poloidal ring limiter configuration, it has been upgraded to a tokamak named ADITYA Upgrade (ADITYA-U) to realize the shaped-plasma operations in an open diverter configurations. The upgradation was conceptualized in the year 2014 and the ADITYA tokamak had been dismantled to its base level in the year 2015. The ADITYA-U tokamak construction was completed in the year 2016 and the first plasma was achieved in December 2016.

The ADITYA-U tokamak is a medium-size air-core tokamak with a major radius of 0.75me and a minor radius (limiter radius) of 0.25 m and maximum toroidal magnetic field of 1.5 T. It has a toroidal belt limiters and two quarter poloidal limiters at two toroidal locations made up of graphite. The ADITYA-U is designed to produce circular plasmas with plasma current ~ 150 – 250 kA, plasma duration of ~ 250 – 350 ms with a chord average electron density in the range of 1 – 5 x 10^19 per meter cube and electron temperature ~ 300 eV – 1000 eV. In addition, it is designed to achieve shaped plasmas with plasma current ~ 100 – 150 kA, elongation (k) ~ 1.1 - 1.2 and triangularity ~ 0.45.

Since 2017, ADITYA-U is routinely being operated and maximum plasma current of ~ 200 kA and longest plasma duration of ~ 350 ms has been achieved in circular plasmas of ADITYA-U tokamak with toroidal belt limiter. Several novel experiments such as electromagnetic particle injection for disruption studies, rotation reversal and saturated Ohmic confinement modes, runaway electron dynamics with gas-puffs etc. are being carried out mainly in circular plasmas. The divertor coils are charged and shaping of the plasma current has been attempted recently obtaining some very encouraging results.

The ADITYA-U tokamak has been operated at the full design value of 1.5 Tesla for the main (toroidal) magnetic field, combined with a long plasma duration of 360 milliseconds, which is 20% higher than its Original design value.

ADITYA-U tokamak has successfully demonstrated full Deuterium plasma operation for the first time in India. The long discharges (~300ms) having ~140 kA of plasma current with deuterium as fuel gas (pre-fill as well as the gas-puffs: all deuterium) in ADITYA-U have been obtained at a main (toroidal) magnetic field of ~1.3 Tesla. This operation also demonstrating an electron density of 3 x1019 m-3. Spectroscopic measurements of the intensities of both D- and H- lines (due to wall recycling as wall is loaded with hydrogen from previous discharges and also from discharge cleaning in hydrogen) was also recorded.

 

[Karthi]

ITER 4.8 KM OF CRYOLINES PASS LEAK TEST

The cryoline network is built out of individual spools that measure up to 10 metres in length and range from 25 to 1000 millimetres in diameter. A section of cryoline can host up to six or seven "process pipes," each devoted to a specific fluid, flow direction or function. The Indian Domestic Agency, responsible for 100 percent of cryoline procurement, is working with two companies—France's Air Liquide and India's INOXCVA. The Indian firm is providing the totality of the piping inside the cryoplant.

One of the first and most obvious requirements for the cryoline network is to be perfectly leak-tight. After more than a full year of work, an important milestone was passed last week as all 4.8 kilometres of cryolines and "warm lines" inside the cryoplant were successfully leak-tested by the INOXCVA team operating under the responsibility of ITER India.

The network inside the cryoplant comprises close to 1,000 spools connected by an equal number of welds. About half the welds were randomly helium-tested prior to the final pressure test at 30 bars, which demonstrated the leak-tightness of the entire network.

 

[Karthi]

Diagnostic Shield Module (DSM) for the Diagnostic ports of ITER tokamak developed in India.

Design of DSM

Final product .This is used in DSM of ITER, DSM is for Shielding against the Neutrons produced by nuclear reaction. these are based on hot pressed boron carbide blocks/pellets.

vacuum barrier developed by India for using in fusion reactors and accelerators to deliver RF power from generators to systems which are working in high vaccum.

 

[Karthi]

CIMPLE -PSI Device , Indias first linear magnetized plasma system for fusion related plasma surface interaction studies.

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Cryocooler based Helium Circulation System at 55K was successfully installed at IPR. The main purpose of this system is to cool the High Temperature Superconductor (HTS) applications (i.e. HTS cables, coils, transformers etc.) at 55K with forced flow cooling method. The SPC-4 Cryogenerator, a Cryocooler which is based on Stirling cycle, is used to generate the low temperature in the system.

The cryogenerator sysytem

Top View of the cryostat

SPC-4 Cryogenenrator with cryostat and cryofans


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Pulsed Plasma Accelerator Laboratory .
Accelerated plasma beam or the plasma jet, that is formed by applying high voltage pulsed, has been attempted for many applications. These includes probable re-fueling into fusion reactor, mitigation of disruption in tokamak or electromagnetic plasma thruster

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Prototype of BSC and HAM for LIGO India . Basic Symmetric Chambers (BSC) are used to support core optical components while Horizontal Access Module (HAM) are used to house auxiliary optics.

3D View of BSC and HAM

 

[Karthi]

Proposed LI-VISTA Test Facility.

Beam tube is one of the important part of the entire vacuum system. It is a cylindrical structure having 4 km length in each perpendicular direction. The beam tube is kept in ultra-high vacuum so that scattering of laser beam is minimized within the Fabry-Perot cavity. Each beam tube has two (02) numbers of beam tube modules of 2.0 km length each. Each beam tube module will be fabricated from 20 m long beam tube sections of 100 numbers. These beam tube sections will be reinforced with vacuum stiffeners as well as support stiffeners. Vacuum stiffeners are provided at a separation of 750 mm distance throughout the length so that the tube can withstand pressure difference between inside and outside the tube and avoid collapsing. Support stiffeners are provided at 20 m apart to stiffen and assemble the support structure for the beam tube.

Beam Tube Support :- Three types of support structures are used. They are (a) fixed supports, (b) guided supports and (c) termination supports are also attached with the support stiffeners as shown in fig (b)


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Indigenously Developed 80 K Sorption Cryopump For SST-1.
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The LLCB TBM set

Schematic and experimental setup for (L) The measurement of effective thermal conductivity of Li2TiO3 pebble bed using Steady State Axial Heat Transfer method (R) The measurement of effective thermal conductivity of Li2TiO3 pebble bed using transient hot wire method

Design and development of Indian Test Blanket Module (TBM) for ITER and future fusion reactors. It has successfully completed the Conceptual Design Review (CDR) for Lead Lithium cooled Ceramic Breeder (LLCB) TBM in September 2015. The TBM structure, being subjected to various load combinations like high temperature and high pressure loads, electromagnetic loads during plasma disruption, and seismic loading conditions, has to demonstrate the structural integrity by carrying out detailed thermal-hydraulic and thermo-mechanical analysis.
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Dusty plasma experimental system

The negative hydrogen ion extraction system and HV power supplies

 

[Karthi]

The ECRH system on Aditya-U

The 42GHz-500kW ECRH system





Electron Cyclotron Resonance Heating (ECRH) system helps tokamak for reliable plasma start-up, heating, current drive and instability control. The ECRH system consists of high power microwave source (Gyrotron), corrugated waveguide based transmission line and mirror based quasi-optical launcher. The Gyrotrons are high power microwave source capable to deliver megawatt level of continuous power at high frequencies varies from 28GHz to 170GHz.

The 42GHz-500kW ECRH system has shown remarkable achievements on the tokamaks SST-1, Aditya and Aditya-U. The Gyrotron delivers 500kW power at -50kV beam voltage, 20A beam current and +20kV anode voltage. This Gyrotron is installed on a cryomagnet which is cooled at liquid helium temperature while carrying out experiments on tokamaks.

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Cryostat for ITER Made in India.

Top lid segments of the cryostat top lid under assembly at ITER-India cryostat workshop in ITER prior to welding

Welding of the segments underway with a robotic arm.

The cryostat is a 30 m tall 30 m diameter large thermos and the biggest vacuum vessel ever built to shield the super conducting magnets and the machine, the cooling water system removes about 500 MW of heat load from the various machine components during operation, the Cryolines and cryo-distribution system is the interface between the cryo-distribution plant and the super conducting magnets and other auxiliaries like the cryopumps to transport croygens and cool to the components to sub-zero temperatures and the in-wall shields sandwiched between the walls of the double walled vacuum vessel act as neutron absorbers.

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In-wall shield blocks supplied by India under installation in the vacuum vessel sectors in Korea

Cooling plant installation nearing completion at ITER site with all components supplied by India