Breeding Blanket Research Articles

Thermal-hydraulic study of the EU-DEMO Helium Cooled Pebble Bed Breeding Blanket Primary Heat Transport System

The European DEMO reactor will work under normal operating conditions in accordance with a pulsed duty cycle. However unplanned and planned transients of plasma overpower may occur compromising the integrity of its plasma-facing components structures. Consequently, adopting appropriate tools is essential to accurately and consistently model the thermal-hydraulic response of the involved cooling systems under both normal operating conditions and accidental events. Given this background, the University of Palermo, in collaboration with EUROfusion, started a research work to study the thermal-hydraulic behaviour of the Primary Heat Transport System (PHTS) of the Helium Cooled Pebble Bed Breeding Blanket (HCPB BB) of the DEMO device under steady-state and transient conditions. The activity has been performed adopting a computational approach, employing the thermal-hydraulic system code TRACE version 5.0 patch 6. The key point of the work has been the code-to-code benchmark with the outcomes previously obtained with the RELAP5-3D code, to estimate the impact of the physical models, numerical resolution schemes and modelling techniques adopted on the predictive capabilities of the system codes considered. The models and the analysis results are presented and critically discussed herein.

Thermal and structural analysis of the European water-cooled lithium lead breeding blanket concept in the L-H mode transition

The design of a breeding blanket system (BB) which reliably fulfils the prescribed requirements is pivotal for the construction of the EU DEMO fusion reactor. For this reason, the BB conceptual design activities require a tighter interaction with plasma physics modelling, so to employ relevant boundary conditions to compare, and then down-select, the BB design alternatives. The preliminary thermo-mechanical study herein reported highlights the importance to assess in deep the BB performance during plasma operational transients. In particular, analyses of the BB thermal and mechanical transients during plasma ramp-up have been carried out, with focus on the L-H transition. Attention has been paid to the Water-Cooled Lithium Lead (WCLL) BB concept, one of the two driver blanket candidates for the EU DEMO project. Adopting DEMO-relevant plasma configurations, the fusion power and the radiative power time evolutions during the plasma ramp-up phase have been modelled under different assumptions using the code ASTRA. The obtained power transients have been used to calculate the time-dependent nuclear power density and radiative heat flux the WCLL BB is subjected to, in order to perform the corresponding transient thermal analysis. Then, the structural analysis has been carried out in order to verify the fulfilment of the RCC-MRx structural design criteria in the most critical time steps. The performed thermo-mechanical analysis has been carried out adopting the Abaqus FEM code. Results obtained are presented and critically discussed, highlighting their importance for the EU DEMO integrated design.

Henry's constant of helium in liquid alkali metals

The determination of the solubility of helium in liquid lithium and in lead-lithium eutectic systems is of crucial importance for the assessment of the reliability, safety and efficiency of future nuclear reactor plants based on fusion technology. In this way, an important aspect to be considered concerns the possibility of accumulations of gas inside the so called breeding blankets of the reactors, due to the low solubility of helium, leading to potentially undesired effects. In the present work we have performed molecular dynamics simulations in order to examine the characteristics of the solubility of helium in alkali metals along a temperature interval that covers near ambient conditions up to temperatures of the order of a 1000 K. Our methodology relied on the classical Kirkwood perturbative approach, which involves the gradual coupling between the infinite diluted He atom and the solvent. Such insertion proceeds along three stages that involved a previous incorporation of a softly interacting cavity that prevented singularities due to the harshly repulsive nature of the actual He-Li potential. Simulated structure factors predicted by the force fields of the different alkali metals were found to be in excellent agreement with available experimental information. Moreover, at all thermal conditions, our results of Henry's constants predict a steady volatility reduction with increasing temperatures, in good agreement with available experimental data. Furthermore, Henry's law volatility constants increase as the sizes of the atoms of the host alkali metals become more comparable to the size of the He atom (the predicted solubility increases with the size of the alkali atoms), and are coherent with the scarce experimental data available.

Neutronics analysis for conceptual design of target system based on a deuteron accelerator-driven fusion neutron source

A Breeding Blanket (BB), producing tritium and heat energy converted to electricity, is a critical component for the successful Fusion Demonstration Reactor (DEMO). A dedicated neutron source is required to validate the performance of the DEMO BB in a continuous long-term fusion-like environment. Therefore, a pre-conceptual design study has been carried out for the Integrated Breeding Test Facility (IBTF) to evaluate the DEMO BB performance. The IBTF is based on a 40 MeV deuteron accelerator-driven system with a maximum current of 10 mA. A solid Beryllium (Be) target with a 20 cm width and height is used to irradiate the tritium breeding unit (TBU).This study conducted a neutronics analysis of the Be target system of the IBTF. The produced neutron energy spectrum and yield were evaluated on a beam energy of 30 to 50 MeV. The analysis results show that the forward direction neutron yield with the 40 MeV deuteron is the highest at 4.92E + 15 n/s, and the neutron ratio of 10–20 MeV is also the highest. A Blistering Mitigation Layer (BML) was added to prevent the blistering effect caused by the deuteron beam within the Be target. A comparison was done for the performance evaluation of the BML candidates vanadium, palladium, tantalum, and niobium. There were no significant differences in the forward direction neutron yield for the BML options. Vanadium was the most beneficial based on the activation analysis results. Therefore, vanadium with a 0.3 mm thickness was added behind the Be target with a 5 mm thickness. In addition, neutronics analysis was conducted for the test cell shielding designs. Concrete 3.5 m thick was required to satisfy a dose of less than 100 uSv/h during the beam operation. When evaluating the dose rate during maintenance, a dose along the cooling pipes was observed due to the activation of the pipes. The neutron flux with the cooling pipes was reduced by using B4C in the maintenance cell shielding, the dose rate was below 100 uSv/h in the test cell.Deuteron and neutron transport calculations were carried out using MCNP6 with the TENDL2019 cross-section library. Activity analysis and gamma source production were performed using FISPACT-Ⅱ with the TENDL2017 cross-section library.

Challenges towards an acceleration in stellarator reactors engineering: The dual coolant lithium–lead breeding blanket helical-axis advanced stellarator case

Under the ambitious EUROfusion mission of “bringing the stellarator line to technological maturity”, the development of a Dual Coolant Lithium–Lead (DCLL) Breeding Blanket (BB) for a Helical-Axis Advanced Stellarator (HELIAS) started. The BB is the crucial systems to achieve tritium self-sufficiency in fusion power plants. The DCLL has potentialities to answer the challenges posed by complex HELIAS configuration, having liquid breeder and decoupled first wall (FW) and breeding zone circuits. Both characteristics are being empowered to simplify remote maintenance and integration of the BB segments. Hence, a “quasi-toroidal” BB segmentation has been proposed to minimize the magnetohydrodynamic pressure drop eschewing complex electrical insulating components while hindering maintenance. Such controversy motivated the use of a detached FW small panels manageable through ports. On this basis, a decoupled Capillary Porous System FW has been assessed under thermal-hydraulics and neutronics point of view. Moreover, ad-hoc tools have been developed for the parametrization of models and convergence toward a viable DCLL HELIAS design, speeding-up the coupling of computer-aided design modelling and analyses. Such novel strategies are described in view of a potential acceleration of the stellarator reactor engineering and by providing an overview of the stellarator challenges inside a global context.

Investigation of the size-dependent crushing behavior of Li4SiO4 pebbles via discrete element method

Li4SiO4 pebbles are commonly employed as breeder materials in solid breeder blankets. As a typical brittle material, the Li4SiO4 pebble sizes significantly affect their crushing behavior. In this study, the size-dependent crushing behavior of pebbles is investigated via discrete element method. Based on the Weibull theory, the probabilistic crushing strength of Li4SiO4 pebbles is analyzed. The numerical results demonstrate that both the characteristic crushing strength and Weibull modulus display a negative linear relationship with Li4SiO4 pebble size in a double-logarithmic plot. Furthermore, by considering both the axial force-displacement responses and fracture process of pebbles, three typical failure patterns of Li4SiO4 pebbles can be identified, namely, splitting failure (Pattern-I), progressive fracturing failure (Pattern-II) and explosive failure (Pattern-III). As the pebble size decreases, the failure pattern gradually shifts from Pattern-I to Pattern-II and to Pattern-III. Additionally, the fragment size distributions of crushed Li4SiO4 pebbles are analyzed based on the fractal theory. It is observed that the fragment size distributions of Li4SiO4 pebbles with varying sizes exhibit favorable fractal features. Smaller Li4SiO4 pebbles exhibit a more significant fragmentation, as evidenced by the increasing fractal dimension of the fragment size distribution with the decreasing pebble size. This work may provide valuable support for further investigations into the mechanical performance of a Li4SiO4 breeder blanket using DEM simulations.

Load study and structural analysis for CFETR divertor OVT

China Fusion Engineering Test Reactor (CFETR) is an important device for studying fusion reactor engineering in China. The Outer Vertical Target (OVT) are high heat load components of the divertors and work in a strong electromagnetic field, so they need sufficient heat removal capability and structural strength. In addition, due to activation of neutrons on materials, the OVT also needs to meet the requirements of Remote Handling (RH) maintenance. CFETR puts forward the concept of hybrid divertor-blanket, which cancels the cassette and adopts a separate structure. The lower part of the divertor is occupied by the breeding blanket, and the divertor target is fixed on the breeding blanket, to improve the RH maintenance efficiency and Tritium Breeding Ratio (TBR). In previous work, the structural design and thermal-hydraulic analysis of the OVT were completed. In this paper, load study and structural analysis was carried out to verify the structural strength of the OVT under experimental test conditions (including liquid pressure, nitrogen baking), normal operation conditions (including the impact of thermal stress under thermal load, gravity, working pressure, etc.) and extreme event conditions (Vertical Displacement Event (VDE) or Major Disruption (MD) event, including the impact of thermal stress under thermal load, gravity, working pressure, electromagnetic load, etc.). The results show that the OVT meets the requirements of heat removal and strength, and effectively promote the processing and manufacturing of the prototype of the OVT.

An Application of a Higher Computational Efficiency Method in Thermo-Hydraulic Calculations of First Wall in Fusion Blanket

An Application of a Higher Computational Efficiency Method in Thermo-Hydraulic Calculations of First Wall in Fusion Blanket

STEAM Experimental Facility: A Step Forward for the Development of the EU DEMO BoP Water Coolant Technology

Within the EUROfusion roadmap for the technological development of the European-DEMOnstration (EU-DEMO) reactor, a key point has been identified in the discontinuous operation (pulse-dwell-pulse) of the machine. Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) adopt technology and components commonly used in nuclear fission power plants, whose performances could be negatively affected by the above mentioned pulsation, as well as by low-load operation in the dwell phase. This makes mandatory a full assessment of the functional feasibility of such components through accurate design and validation. For this purpose, ENEA Experimental Engineering Division at Brasimone R.C. aims at realizing STEAM, a water operated facility forming part of the multipurpose experimental infrastructure Water cooled lithium lead -thermal-HYDRAulic (W-HYDRA), conceived to investigate the water technologies applied to the DEMO BB and Balance of Plant systems and components. The experimental validation has the two main objectives of reproducing the DEMO operational phases by means of steady-state and transient tests, as well as performing dedicated tests on the steam generator aiming at demonstrating its ability to perform as intended during the power phases of the machine. STEAM is mainly composed of primary and secondary water systems reproducing the thermodynamic conditions of the DEMO WCLL BB PHTS and power conversion system, respectively. The significance of the STEAM facility resides in its capacity to amass experimental data relevant for the advancement of fusion-related technologies. This capability is attributable to the comprehensive array of instruments with which the facility will be equipped and whose strategic location is described in this work. The operational phases of the STEAM facility at different power levels are presented, according to the requirements of the experiments. Furthermore, a preliminary analysis for the definition of the control strategy for the OTSG mock-up was performed. In particular, two different control strategies were identified and tested, both keeping the primary mass flow constant and regulating the feedwater mass flow to follow a temperature set-point in the primary loop. The obtained numerical results yielded preliminary feedback on the regulation capability of the DEMO steam generator mock-up during pulsed operation, showing that no relevant overtemperature jeopardized the facility integrity, thanks to the high system responsivity to rapid load variations.

Numerical simulation of tritium extraction from liquid PbLi by gas–liquid contactor

As one key component of the fusion reactor, the blanket is in charge of breeding tritium, shielding neutrons and extracting thermal energy for electricity generation. An efficient tritium extraction unit (TEU) is a necessary guarantee for the tritium self-sufficiency in the fusion reactor. The gas–liquid contactor (GLC) is considered to be one of the most efficient devices for the liquid blanket that uses the lead–lithium (PbLi) as the functional materials. Understanding the characteristics of gas–liquid two-phase flow and mass transfer is beneficial for guiding the design and optimization of contactor. However, it is difficult to carry out the tritium experiments due to the scarcity in nature, the strong permeability, as well as the radioactivity. In this paper, the turbulent two-phase model coupled with dilute mass transfer is adopted to numerically study the process of tritium extraction from the liquid PbLi, and the numerical model is evolving step-by-step. Firstly, based on the models of single bubble and helium purge, the effects of different design parameters on the tritium extraction efficiency of the contactor are further investigated (i.e. mass flow rate of helium, initial tritium concentration and liquid PbLi filling height). The results are in good agreement with previous experimental results, and this validates the feasibility of the numerical model used in this work. Then, the entire model of GLC is numerically analyzed, which illustrates that the tritium extraction efficiency of one single unit is only 20.8%. However, the total efficiency can achieve 95% when there are 13 units working together in series. On this basis, the schematic diagram of tritium extraction system is proposed for the fusion blanket.

The Design of Water Loop Facility for Supporting the WCLL Breeding Blanket Technology and Safety

The WCLL Breeding Blanket of DEMO and the Test Blanket Module (TBM) of ITER require accurate R&D activities, i.e., concept validation at a relevant scale and safety demonstrations. In view of this, the strategic objective of the Water Loop (WL) facility, belonging to the W-HYDRA experimental platform planned at C.R. Brasimone of ENEA, is twofold: to conduct R&D activities for the WCLL BB to validate design performances and to increase the technical maturity level for selection and validation phases, as well as to support the ITER WCLL Test Blanket System program. Basically, the Water Loop facility will have the capability to investigate the design features and performances of scaled-down or portions of breeding blanket components, as well as full-scale TBM mock-ups. It is a large-/medium-scale water coolant plant that will provide water coolant at high pressure and temperature. It is composed by single-phase primary (designed at 18.5 MPa and 350 °C) and secondary (designed at 2.5 MPa and 220 °C) systems thermally connected with a two-phase tertiary loop acting as an ultimate heat sink (designed at 6 bar and 80 °C). The primary loop has two main sources of power: an electrical heater up to about 1 MWe, installed in the cold side, downstream of the pump and upstream of the test section, and an electron beam gun acting as a heat flux generator. The WL has unique features and is designed as a multi-purpose facility capable of being coupled with the LIFUS5/Mod4 facility to study PbLi/water reaction at a large scale. This paper presents the status of the Water Loop facility, highlighting objectives, design features, and the analyses performed.

PbLi/Water Reaction: Experimental Campaign and Modeling Advancements in WPBB EUROfusion Project

The Water-Cooled Lithium–Lead blanket concept is a candidate breeding blanket concept for the EU DEMO reactor and it is going to be tested as one of the Test Blanket Modules (TBM) inside the ITER reactor. A major safety issue for its design is the interaction between PbLi and water caused by a tube rupture in the breeding zone, the so-called in-box LOCA (Loss of Coolant Accident) scenario. This issue has been investigated in the framework of FP8 EUROfusion Project Horizon 2020 and is currently ongoing in FP9 EUROfusion Horizon Europe, defining a strategy for addressing and solving WCLL in-box LOCA. This paper discusses the efforts pursued in recent years to deal with this key safety issue, providing a general view of the approach, a timeline, research and development, and experimental activities. These are conducted to master dominant phenomena and processes relevant to safety aspects during the postulated accident, to enhance the predictive capability and reliability of selected numerical tools, and to validate and qualify models and codes and the procedures for their applications, including coupling and chains of codes.

Characterization of aged Li4SiO4 pebbles performance

Many studies and research have been carried out in the recent years to identify which breeder material and (in which) form, solid or liquid, are most suitable to be used in the Breeder Blanket (BB) modules. Nonetheless, the choice of a stable material, with high thermal conductivity and melting point remains unsolved and becomes even more complex when durability is considered. The breeder material must guarantee that the thermo-mechanical properties change as little as possible under environmental and operating conditions. The aim of this study is to investigate the ageing effects on the mechanical behaviour of lithium orthosilicate (Li4SiO4) in form of pebbles. The pebbles had been produced at the University of Pisa at room temperature by a hydrolytic sol-gel supported drip casting method and subsequently stored for 4 years in an environment simulating normal storage (i.e., standard condition). These pebbles of about 1.2–1.4 mm diameter have been subjected to uniaxial compression tests, without radial constraints, and SEM analysis. The results obtained were compared to those of non-aged pebbles to highlight the degradation caused by the ageing. SEM examination of aged pebbles showed increased porosity and a pebble surface that was not entirely homogeneous and uniform. This affected the crack shapes on the contact surface and the crushing load.

Magnetohydraulic flow in a rectangular channel filled with stream-wise obstacles

The Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB) is one of the two leading candidates for implementation as driver blanket in the European DEMO reactor, which is expected to start operation in the late 2050s and serve as a BB test facility. One of the main risks associated with the current WCLL concept is the large number of welds necessary to realize its horizontal (toroidal-radial) breeding zone cooling system and stiffening structure, which negatively affects its reliability. Hence, an alternative variant concept has been proposed which relies on vertical (radial-poloidal) “double-bundle” cooling tubes and stiffening plates: the WCLL double-bundle (WCLL-db). In this paper, we report the results of numerical analyses performed to characterize the Magnetohydrodynamics (MHD) regime in the WCLL-db where the liquid metal breeder flows vertically and it is obstructed by a large number of electrically conductive obstacles aligned with the stream-wise direction.Direct numerical simulations are performed with the aid of computational MHD tools to characterize the flow features and pressure losses in the WCLL-db breeding zone. The presence of many stream-wise obstacles affects the velocity distribution since the cooling tubes provide additional paths for the electric current closure. The most striking feature is the breaking down of the classic MHD slug core flow into many separate smaller regions, separated by internal layers. These occur tangential to the pipe walls, featuring velocity overshoots, and tend to propagate along the imposed magnetic field. A moderate (3.83–9.73%) pressure penalty is found compared with the analytical prediction for a channel devoid of obstacles.

Reactor experiments on irradiation of two-phase lithium ceramics Li2TiO3/Li4SiO4 of various ratios

The design of future fusion reactors involves the production of tritium inside the breeder blanket. The most promising material for solid breeder blankets is a two-phase lithium ceramic containing orthosilicate Li4SiO4 (LOS) and metatitanate Li2TiO3 (LMT) of lithium in various proportions. Tritium is formed in lithium under neutron irradiation by the reaction 6Li(n,α)T. Further, this tritium is extracted from the blanket with a purge gas and returned to the fusion zone, realizing the concept of a closed fusion cycle.Irradiation under fission reactor conditions is still one of the few available methods for estimating the parameters of tritium generation and release from lithium-containing materials in the "in-situ" mode. This paper presents the results of experiments on neutron irradiation of two-phase lithium ceramics of various ratios (LOS + 35 mol. % LMT (pebble size 250–1250 μm), LOS + 35 mol. % LMT (pebble size 500–710 μm) and LOS + 25 mol. % LMT (pebble size 500–710 μm) at the WWR-K research reactor.Irradiation of each batch of samples lasted from 5 to 22 days. The experiments were carried out by the vacuum extraction method. This paper describes the main methodological aspects of the studies, namely the technical features of four irradiation campaigns, the sequence and scope of the studies. A comparison is also made of the initial sections of reactor experiments for all campaigns, where the reactor was sequentially brought to power, according to which the parameters of the Arrhenius dependence of the effective tritium diffusion coefficients were estimated.

Calculation of Temperature Fields in a Lithium Ceramic Pebble Bed during Reactor Irradiation in a Vacuum.

Two-phase lithium ceramic Li2TiO3-Li4SiO4 is considered as a tritium multiplier for use in the solid blanket of fusion reactors. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials under neutron irradiation can only be obtained from experiments in fission research reactors. At that, irradiations in vacuum give the possibility to register even very fast gas release processes (bursts) from the ceramics' voids and pores, although it reduces the thermal conductivity of the pebble bed. The purpose of this work was to simulate the heating of mono-sized pebble bed (1 mm in diameter) of two-phase lithium ceramic 25 mol%Li2TiO3+75 mol%Li4SiO4 in an ampoule device during neutron irradiation at the WWR-K research reactor under vacuum conditions, and to determine experimental parameters in order to prevent heating of the lithium ceramics up to the Li4SiO4-Li2SiO3 phase transition temperatures (>900 °C). For the first time, it was obtained that the effective thermal conductivity of a 1 mm mono-sized pebble bed of 25 mol%Li2TiO3+75 mol%Li4SiO4 significantly decreases (four times) when it is irradiated with neutrons in a vacuum (at a helium pressure of approximately 10 Pa), compared to a similar calculation at 100 kPa of helium (when the He sweep is used). It was concluded that it is difficult to evaluate the maximal temperature of the ceramics in the capsule by measuring the temperature of its outer metal wall (according to thermocouple readings) without using the results of thermophysical calculations for each type of ceramic, taking into account its quantity, specific heat release and pebble size(s). To control the temperature of the ceramics during an irradiation experiment in a vacuum, an in-capsule thermocouple should be used, placed in the center of the pebble bed. Measuring the temperature of the pebble bed based on the capsule wall temperature can lead to overheating of the ceramics and phase changes.

Testing needs for the development and qualification of a breeding blanket for DEMO

The purpose of this paper is to critically review the testing and qualification strategy for the DEMO breeding blanket in Europe, identifying the potential risks and weaknesses and recommending, where necessary, the changes required to strengthen or accelerate the programme. Based on the information presented in this paper, a risk mitigation strategy is required to reduce the technological, licencing, and regulatory risks associated with the performance and reliability of the breeding blanket technologies for DEMO. An attractive mitigation option, which was originally proposed more than 30 years ago but not implemented, is to perform testing and qualification of breeding blanket technologies and design concepts in a dedicated nuclear plasma device that serves as a 14 MeV volumetric neutron source.

Electromagnetic load evaluating and optimizing approach of the blanket system during VDEs considering halo current

In most scenarios, the vertical displacement event (VDE) represents the most extreme electromagnetic (EM) event within the tokamak device. The significant EM loads experienced during this event have the potential to compromise the structural stability of in-vessel components. This study investigates the EM loads on the water-cooled ceramic breeder blanket system of China Fusion Engineering Test Reactor (CFETR) using finite element analysis methods in two characteristic events: hot-VDE and cold-VDE. The study discusses the EM load effects resulting from changes in magnetic flux and induced electromotive force, respectively, with a specific focus on halo currents. The results reveal that, with similar current quech time, the difference in EM load on the blanket system during the VDEs primarily depends on the halo currents. When the electrical connection of the back supporting structure (BSS) is open, the halo current path within the blanket system and vacuum vessel (VV) changes, and a substantial portion of the halo current in the blanket system is conducted to the VV via the BSS. Consequently, a portion of the EM load on the blankets and BSS is transferred to the VV due to the transfer of halo current. Inspired by this, the conceptual use of ‘shunts’ is proposed to provide a dedicated circuit for shunting halo currents away from critical device components, such as the VV and blankets. This approach allows for the sharing of EM loads caused by halo currents and reduces the threat posed by halo currents to the structural integrity of these essential components.

The effect of blanket system and port plugs on the harmonic components of error field and optimizing approach investigation

The fusion plasma is sensitive to the penetration of the toroidal mode numbers harmonic components of the error fields (EFs), and the toroidal mode components of the EFs are closely tied to the toroidal distribution of the magnetic field source. The primary source of EFs within the magnetic field sources is the current source, such as the magnet system, while the magnetic field produced by magnetized ferromagnetic materials may also generate EFs. The ferromagnetic field source, such the reduced-activation ferritic-martensitic steel, represent a smaller portion of the EFs source. However, the use of ferromagnetic materials is expected to increase significantly as tokamak research transitions from experimental devices to demonstration and commercial reactors, particularly for the structural materials of in-vessel components and port plugs. Predictably, the EFs induced by ferromagnetic materials will increase and may even surpass the design specifications. In this study, the EFs introduced by the water-cooling ceramic breeder blanket system and the vacuum vessel port plugs of China Fusion Engineering Test Reactor are comprehensively analyzed by an analytical approach and finite element numerical approach, which is based on the Fourier decomposition. The results reveal a clear linear relationship between the EFs formed by the saturated magnetization module and its rigid displacement deviating from periodic symmetry, as well as the same behavior in EFs and overall magnetization intensity. Compared to the assembly error of the blanket system, the EFs that may be introduced by the equatorial ports and plugs are substantial, including larger toroidal mode harmonic components that can lead to lock mode and disruption. Moreover, using the correlation between harmonic components and symmetrical periods, specific toroidal mode components can be selectively shielded.

Overview of the Experimental Helium Cooling (EHCL) system

The objective of the Experimental Helium Cooling (EHCL) facility at IPR is to contribute to the development of the helium cooled nuclear fusion components such as tritium breeding blanket and divertor by providing experimental data on the mock-up testing of these components. All these mock-ups have several complex cooling circuits created for efficient heat extraction by high pressure helium gas. Several such mock-ups to be tested beforehand for evaluating the thermal performance and concluding the design & fabrication techniques of tritium breeding blankets and divertors. Moreover, studying performance of the integrated plant, learning behaviour of different EHCL components e.g. circulators, heat exchangers, heaters, compressors & valves etc., and demonstration of operational & control scheme under different operating conditions are essential for development of the helium cooling system (HCS) for Indian Test Blanket Module (TBM) for ITER. This paper describes the EHCL facility that is being commissioned at IPR. It briefly presents results of the thermo-hydraulic studies of the integrated system. It also talks about its major components and their performance results, operation and control scheme of the loop, and highlights the key features of the data acquisition & control system of the facility.