Fusion & Related Technologies
Fusion & Related Technologies
To realise thermonuclear fusion, a vast majority areas of different technologies have to work in tandem: High current electromagnets, advanced material technologies, robotic technologies, cryogenic technologies, beam technologies, Radio-Frequency wave technologies etc. Also these technologes have to work in a very hostile environments which includes nuclear radiations. Continuous progress related to these technologies are being made with relevant science and engineering.
Magnet technology is one of the important requirements for the magnetically confined tokamak plasmas, as it takes the major share of the fusion machine. For steady state operations superconducting magnets are preferred over copper magnets. Starting from Cryogenic temperature based superconducting magnets, efforts are being carried out to apply high temperature superconducting magnets.
Fabrication and testing of compact HTS solenoid coil: High temperature superconductors (HTS) are promising candi-dates for next generation high field compact magnets, since they do not require cryogenic cooling at 4 K. Coil winding, inter-pancake and terminal joints are challenging technologies required for the fabrication of HTS tape based high-field magnets. As a first step, IPR has fabricated a small bore (~50 mm) double pancake, liquid nitrogen bath-cooled, HTS solenoid coil with 24 turns and a height of 21 mm. This has produced a pulsed magnetic field of 1.1 Tesla with a current pulse of duration 0.72 millisec, and a DC magnetic field of 0.06 Tesla at 110 A current. Figure A.4.1.1 shows the HTS coil assembly used to measured magnetic fields. The next step will be to develop higher-field and larger bore HTS coils for practical applications these activities are covered in a new DPR titled “Fusion Technologies” that has been cleared by PAC and is awaiting DAE sanction.
Figure A.4.1.1 HTS coil assembly
Figure A.4.1.2 Hybrid Joint integrated with the test insert
Indigenous Development of Hybrid Nb3Sn and NbTi CICC joint: Systems with superconducting magnets often require joints between different kinds of superconducting materials. For the first time in India, IPR has developed a hybrid over-lap joint of length 120 mm, for connecting Nb3Sn and NbTi Cable-in-Conduit-Conductor (CICC). This is a thermally stable joint operating at 4.5 K and currents upto 10 kA. This joint is of practical importance because it can be integrated in the limited space available near Nb3Sn CICC-based super-conducting magnets
Integrated Magnet Test Facility:An integrated magnet test facility (MTF) has been installed & commissioned at IPR. This facility is meant for testing high temperature (80-4.5 Kelvin) as well as low temperature (~4.5 Kelvin) supercon-ducting magnets for fusion & other applications. It consists of a large Cryostat weighing 21 Tons and having a volume of 83 m3, along with a liquid nitrogen thermal shield. It is capable of housing superconducting magnets with a maxi-mum size of 5 m height and 4 m diameter. The vacuum and cool down performance of the thermal shields of this cryostat have been found to be satisfactory.
Figure A.4.1.3. (a) Integrated Magnet Test Facility (MTF)
(b)Thermal shields of bottom part of cryostat
of the MTF
Plasma Facing Components Technologies
The materials facing the hot plasmas are in extreme conditions (neutrons+heat). Therefore these are critical to the success of a fusion reactor. The materials facing the fusion plasma will have to face 14 MeV neutrons as well as very high temperatures (5 million degree C). This makes the matter to get eroded/ablated. The design of a fusion reactor thus calls for a layer of plasma facing materials which can not only sustain the environmental conditions in the core but also becomes helpful in extracting heat. The blankets are located behind the plasma facing comonents.
Clockwise from top: The HHFTF system showing (A) the electron beam chamber and (B) the water-cooling system. (C) Closeup of the electron beam gun of the system
The High Heat Flux Test Facilty (HHFTF) is state-of-the-art facility for material plasma interaction studies which is available in very few places around the world. This helps to decide the armour materials for fusion machines which needs to withstand the extreme conditions prevailing in the edge of the tokamak
i) A new copper-alloy test mock-up of 400 mm length has been specially developed for conducting experimental stud-ies on Critical Heat Flux(CHF) for One-Sided Heating using the High Heat Flux Test Facility (HHFTF). Critical Heat Flux for quasi-steady heat flux condition is observed during the experiments. Subsequent experiments with transient heat flux indicate that CHF for fast rising transient heat-flux is lower than the quasi-steady heat flux.
ii) HHFTF has been used to perform heat removal testing on the Back-Plate of Positive Ion Neutral Beam Injector (PINI). Electron Beam of HHFTF was operated at full power of 200kW at 45 kV acceleration voltage for ~450 s to generate a steady-state heat flux of 2.5 MW/m2 on the entire surface of the back-plate. Infra-Red Mapping of the surface of back-plate is performed to generate 2D temperature profile during High Heat Flux testing.
iii) A Tungsten-coated target (450 mm long, 30 mm wide) with copper-alloy substrate, developed by ARCI, has been successfully tested for its performance at 500 C surface temperature (3MW/m2 heat flux) for 1000 thermal cycles using HHFTF. Tungsten coated sample is also successfully tested during thermal fatigue tests performed using the Gleeble system for 1000 cycles from 100 to 500oC.
iv) Brazed Tungsten Mono-block fabrication technique de-velopments/improvements for divertor targets are continued. Improved machining techniques resulted in achieving better concentricity of machined hole in tungsten and casted copper within +/- 100 microns as per ultrasound investigations. In the attempts made to braze copper-alloy tube with copper-casted tungsten mono-block tile using horizontal assembly, brazing is achieved with minor defects.
v) Vacuum Brazing has been performed for development of Straight Section and Y-Section Wave-Guides.
Fusion Blanket Technologies
High energy neutrons generated from the fusion reactor needs to be utilized for breeding tritium fuel as well as for utilizing the energy. The Blanket acts like heat exchanger as well as to generate tritium fuel from lithium. This development work is being currently undertaken at IPR.
Thermal-hydraulics and structural analyses of LLCB TBM set: Indian Lead-Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) is one of the tritium breeding Fusion blanket concepts. First wall (FW) is a plasma facing component of the TBM designed to withstand high heat flux from plasma. FW is cooled by high pressure, high temperature helium gas flowing through coolant channels. The detailed thermal–hydraulics of FW has been performed based on the heat flux and neutronic heat generation. using ANSYS CFX. The distribution of flow in different flow circuits of FW from manifolds has been performed and the flow distribution is found to be uniform in all the circuits. Thermal-hydraulic analysis of helium flow inside the FW channels has been done and then the manifolds to estimate temperature, pressure drop and heat transfer coefficient have been documented. TBM shield is located behind TBM to provide shielding from high energy neutrons to magnets and other components behind the TBM set. Water flows in parallel channels inside TBM Shield, provides the function of neutron moderator as well as coolant to remove heat deposited by the neutrons in the structure. Flow rate in parallel channels of shield has been regulated using orifice of different diameter. CFD Flow analysis inside shield has been performed to validate the distribu-tion of flow inside the parallel channels. Based on velocities obtained, the heat transfer coefficients have been evaluated and thermal analysis of TBM shield has been performed considering thermal load from neutronic heat generation. The results obtained from thermal–hydraulic analysis of FW, manifolds and TBM shield have been used for thermo-structural analysis of LLCB TBM set based on load combinations as per ITER load specifications. RCC-MR 2007 code has been used for the structural assessment and prevention of ptype and s-type damages. The temperature and stresses are found to be within the acceptable limits and safety margins with some proposed modifications.
Performance Assessment of the Helium Cooled First Wall Mock-Up in HELOKA Facility: The Lead-Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) being developed by India for testing in ITER adopts various well developed design engineering and manufacturing technologies. The First Wall (FW), which is directly exposed to the incident heat flux, is designed for high pressure helium flow, high operating temperature (up to 100 bar and 550 °C) and considerable thermal stress and electromagnetic disruption loads. In order to check the thermal performance of the FW and ensure its structural integrity, a mock-up of the FW fabricated in India was tested in HELOKA test facility at KIT, Germany. Both normal and accidental operating conditions were investigated under ITER-like surface heat fluxes. Based on these results the thermal performance of the FW were validated.
3D Modelling of Loop Layout, Pipe Stress Analysis and Structural Responses of High-Pressure High-Temperature Experimental Helium Cooling Loop (EHCL): An Experimental Helium Cooling Loop (EHCL) is being developed as a part of R&D activities in fusion blanket technologies. This system is designed to test various nuclear fusion blanket mock-ups. The primary loop is designed to remove 75 kW heat load on the Test Section Module (TSM). This system is a high-pressure high-temperature loop which produces significant deflections and thermal expansions in the piping network, which leads to the reaction forces and moments. During the earthquake, an additional high acceleration acts (in all directions) on the piping system which again enhances the pipe deflections. EHCL equipment arrangement, loop layout, methodology and results of pipe stress analysis have been studied. EHCL equipment are connected through DN 50 schedule 80 major pipes and associated valves. The high temperature piping network is analyzed for sustained and occasional load responses to ensure the integrity of the system. The process piping code ASME B31.3 is referred for pipe stress analysis. The calculated stresses are in acceptable limit. The least available stress margin is ~29% and the corresponding displacements are of 9.8 mm, 19.72 mm and 21.76 mm in x, y and z directions respectively are observed in the heater outlet to TSM inlet line. The obtained results of reaction forces and moment forces would be utilized as an input for the selection of pipe. supports. The results of pipe stress analysis would be used in further loop optimization.
Automation and Interlock System Design for BP-Li Liq-uid-Metal Purification Experimental Facility: In nuclear fusion domain, eutectic lead-lithium (Pb-Li) is of critical importance as a tritium breeder, neutron multiplier as well as a high-temperature coolant for nuclear power plants. To achieve intended plant efficiency through operation of such high-temperature fluid systems, it is necessary to control composition characteristics of operating fluid. In contrast, at elevated temperatures and higher velocities lying within op-erational region of interest, Pb-Li plays role of an active corrodent for structural materials, exhibiting selective leaching towards certain elements like Ni, Cr, Mn etc. Such elemental impurities alongwith oxides tend to precipitate in cooler sections of the system, thereby restricting flow of coolant resulting in overall performance degradation. To maintain the pu-rity of Pb-Li, online removal of impurities is essential. In this regard, an online Pb-Li purification facility is under fabrication. Performance evaluation of this facility will be assessed over long duration operations at relevant process conditions. To operate the facility with minimum human-intervention and downtime, a PXI-express platform based data-acquisi-tion and control system is designed using LabVIEW environment for continuous monitoring and control of process pa-rameters. Investment protection for critical loop components is addressed using software based interlock modules. The details about the developed data-acquisition system, definitions for control and interlocks logics, alarm management, remote operation of loop components and salient integrated features for a user-convenient long-duration automated operation of the facility are documented. Detailed design and validated performances for cover gas pressure control system and timer-based latched configuration liquid-metal drain in-terlock are also studied.
Development of an Optimised Magnetic Field Source for Flowmeter Applications: Sensitivity of a magnetic flowmeter relies on many factors like magnetic field strength, gap between electrodes, material properties, magnet temperature etc. For a given measuring conditions, a strong magnetic field source can produce highest sensitivity for the flowmeter. An economic design of magnetic field source would be to produce the strongest magnetic field from a given amount of magnetic material. For this purpose, various magnet configurations are analyzed using FEM and the flowmeter sensitivity using such magnet configurations are compared. It is observed that magnets arranged in a Halbach fashion produce the highest sensitivity for the flowmeter using a given amount of magnetic material. The major challenge for the development of such a magnetic field source is its fabrication from its constituent magnets, combatting their huge attractive/repulsive forces (~2500 N for our case). Therefore, a specific mechanical tool has been designed for assembling the magnetic field source and a robust assembly technique has been devised using nu-merical computations. The designed magnetic field source produces a peak magnetic field of 0.78 T in the pole cross section of 50 mm × 50 mm.
A Neutronic Experiment to Support the Design of an Indian TBM Shield Module for ITER: A shield module is associated with an Indian Test Blanket Module (TBM) in ITER to limit the radiation doses in port inter-space areas. The shield module is made of stainless steel plates and water channels. It is identified as an important component for radiation protection because of its radiation exposure control functionality. The radiation protection classification leads to more assurance of the component design. In order to validate and verify the design of the shield module, a neutronic laboratory-scale experiment is designed and executed. The experiment is planned by considering the irradiation under a neutron source of 14 MeV and yields of 1010 n s−1 . The reference neutron spectrum of the ITER TBM shield module has been achieved through optimization of the neutron source spectrum by a combination of steel and lead materials. The neutron spectrum and flux are measured using a multiple foil activation technique and neutron dose-rate meter LB 6411 (He-3 proton recoil counter with polyethylene), respectively. The neutronic design simulation is assessed using MCNP5 and FENDL 2.1 crosssection data.
Remote Handling & Robotics Technology
Robotics is a very important part if fusion technology since inspection, repairs inside the vacuum vessel is not possible as human access is not allowed inside a fusion device due to the radioactivity. For this, development of inspection and robotic arms, tools for specific purposes, Virtual Reality for training manpower on carrying out inspections/repairs etc. are being developed.
The Large format 3D virtuality reality facility at IPR
Development of Robotics at IPR (L) ARIA Arm (Payload: 25kg; Reach: 2.2m) (M) Hy-RIS Snake Arm (R) PRAS (Payload: 5kg; Reach: 1.2m)
The Articulated Robotic Inspection Arm (ARIA) has been designed and developed by IPR for use as inspection arms inside fusion devices. Another novel inspection arm is the Hyper Redundant Inspection System (Hy-RIS), in which the actuators of the arm are located remotely and the control is done using stainless steel cables. This type of inspection arm can be extremely useful for inspections in constrained spaces and inspections though narrow gaps. Since there are no motors of gears on the arm, they can be used in environments where motor is to be isolated from working environment, like in vacuum or high temperatures.
An advanced facility for virtual reality has been established at IPR. This 3-sided, fully immersive integrated Virtual Reality CAVE facility is capable of rendering designs in 3D VR. This facility is seamlessly compatible with various design and modelling software, viz., CATIA v5, Solidworks and 3DVIA composer. The users can load the 3D models of any machines/systems in the VR facility, instantly view these models as 3D immersive models using the stereoscopic glasses, and feel as if they are actually present in the actual environment. The facility has in-built head/hand tracking and the haptic arm for navigation and interaction within the virtual environment. Various features include virtual assembly, collision detection, animations, navigation, zoom in/out, fly through, cutting planes, and snapshots, etc. VR is extremely useful in remote handling as remote operations require an accurate perception of a dynamic environment. The facility gives the operators the same unrestricted knowledge of the task scene as would be available if they were physically located in the remote environment.
Neutral beam is used for heating and as current drive in a magnetically confined fusion power plant. In this, initially ion beams are accelerated to high energy and then neutralized (to avoid the magnetic field) and then injected inside the core vessel. The major challenge lies in neutralizing the accelerated ion beams without losing its energy. Technology related to (i) ion source (ii) acceleration (iii) neutralizers (iv) duct (interface between the plasma and neutral beam system) and (v) beam dump are being developed at IPR.
High power Neutral Beam Injection (NBI) systems are used world wide for fusion reactor plasmas. The heart of an NBI system is the ion source and one of its critical components is a water-cooled Back Plate (BP). The BP, used for mounting several critical components, must also handle a high heat load of 2 MW/m2 from the plasma. This implies considerable technical challenges in fabrication. Following considerable R&D, such a BP has been fabricated for the first time in India and successfully tested on IPR’s High Heat Flux Test Facility (HHFTF) upto 2.5 MW/m2. Hence this critical component can now be manufactured in India at a lower cost.
Figure: Water cooled back plate for NBI source
Performance Evaluation of Various Diagnostics Developed for a Negative Ion Based Neutral Beam Injector Program: The characteristics of a negative hydrogen ion (H-) source and its neutralization efficiency determine the performance of a negative ion based neutral beam injector (NNBI). There-fore, for the safe operation of an NNBI system, it is necessary to monitor the performance of the ion source and its beam through a systematic characterization process. A judicious selection of different diagnostics based on electrical, optical and thermal types and including calorimetric techniques are required. In this regard, a number of diagnostics are being developed under the NNBI R&D program in the Indian Test Facility (INTF). These diagnostics are versatile in nature in terms of their working principles and independent prototype experimental efforts have been carried out to establish them and prepare them for operational use. Electrical probes (EP), optical emission spectroscopy and cavity ring down spectroscopy (CRDS) are mainly envisaged for ion source plasma characterization. Additionally, standard electrical measurements in RF and DC power supply circuits are already in regular use in the operational experimental setups, ROBIN and HELEN-I, for monitoring the power supplies. Doppler shift spectroscopy (DSS) and optical emission to-mography (TOMO) are being developed for beam characterization in terms of divergence, stripping and beam profile. Some of these are characterized on separate prototype experiments and are already integrated and have been tested in the available operational plasma experimental setups: ROBIN and HELEN-I. The DSS system, with multiple lines of sight (LOS) (blue-shifted and red-shifted), is integrated in the ROBIN setup and CRDS is arranged in HELEN-I. The TOMO technique is used to find the beam power density profile from the hydrogen beam emitted Balmer line intensity. The optical brightness profile of a neutral beam due to beam emission radiation is proportional to the beam power density. In this regard, a tomography code based on maximum entropy is developed to reconstruct the 2D optical emissivity profile of the INTF beam by inverting the LOS integral of the brightness of the beam. The code has been validated with the simulated INTF beam power density profile, in terms of the mathematical functions representing it.
Large Volume Cryoplant Systems
The future fusion based power stations are expected to have very large super-conducting magnets which would require very large cryogenic plants. This activity involves building modular cryogenic plants each of 1MW capacity.
Indigenous helium compressor for cryoplant application: One of the major components in a cryoplant is a helium compressor, which is normally imported. As a significant step towards indigenisation, an industry-scale air compressor has been successfully converted to a helium compressor. This has been operated successfully in a closed loop continuously for 24 hrs with a helium flow rate of 60 g/s & delivery pressure of 14.5 bar, with oil impurities in the compressed helium below 100 PPB (parts per billion) and a local leak rate ~10-5 mbar litres/sec. The cost of such a compressor is several times lower than that of imported ones.
Figure: Helium compressor converted from air compressor
- Magnet Technologies
- Plasma Facing Components Technologies
- Fusion Blanket Technologies
- Remote Handling & Robotics Technology
- Beam Technologies
- Large Volume Cryoplant Systems