Thermonuclear Experimental Reactor Research Articles

Engineering design of the CFETR machine

The China Fusion Engineering Test Reactor (CFETR) is the next step in China's roadmap for magnetic confinement fusion development. It aims to bridge the gap between the ITER fusion experimental reactor and the demonstration fusion power reactor (DEMO). The CFETR will be operated in two phases. Phase I aims to achieve fusion power of 50–200 MW, Qplasma = 1–5, TBR > 1.0 and neutron irradiation effects ∼ 10 dpa. Phase II aims to perform tokamak DEMO validation under fusion power > 1 GW, neutron irradiation effects ∼ 50 dpa and figQplasma > 10. To ensure that the CFETR machine can achieve these goals, the major/minor radii of plasma have been altered from R = 5.7 m/a = 1.6 m to R = 7.2 m/a = 2.2 m. The new machine can meet the targets of both Phase I and Phase II and achieve smooth transfer between them. The configuration of the CFETR machine is presented in this paper. Then, the design progress and status of some key components are described, including the magnet system, blanket system, diverter system, remote handling system, vacuum vessel, thermal shield and cryostat. The research and development of most components are underway and will be demonstrated in the near future.

Conceptual Design Studies of an HTS Insert for the DTT Central Solenoid

The “Divertor Tokamak Test” (DTT) facility is an experimental fusion reactor being built in Frascati (IT) in the framework of the European Fusion Roadmap. The DTT Central Solenoid comprises six independently energized modules, wound from Nb3Sn Cable-In-Conduit Conductors (CICC). The future addition of a High Temperature Superconductor (HTS) insert would enhance the CS performance, in terms of magnetic flux density and magnetic field produced in the machine and would represent a fundamental technology demonstration toward high-field fusion. Along with the LTS cables, the Central Solenoid complex aims to reach 18 Wb at the pre-magnetization phase, being 16.54 Wb with the LTS CS only. Each LTS module is internally divided into three sub-modules: High Field (HF), Medium Field (MF) and Low Field (LF), each with a dedicated CICC cable optimized for the intense electromagnetic loads during the plasma scenarios. The LTS cables adopt a 316LN steel jacket and a fiberglass-reinforced resin insulation layer. The conceptual design presented here concerns an HTS insert, to be possibly placed inside the CS bore in the machine, employing a Round-In-Square (RIS) configuration of the conductor with either aluminum alloy or steel jackets. The study carried out for the design of the HTS insert and its pre-compression structure, the main electromagnetic and structural analyses and the design solution are presented in this work.

Experimental Investigation of Tritium Compatible Stainless Steel Roots Pumps

SummaryThe tritium compatible Roots pump Okta 1500 (Pfeiffer Vacuum) is designed for pumping light gas species consisting of diluted radioactive gas, especially tritium mixed with other hydrogen isotopes and helium. Therefore, the pump is constructed of stainless steel and equipped with a special gas lubricated mechanical seal.The series of experiments presented in this work include the evaluation of the performance and an integrity check of the glide ring seal as well as the determination of the light gas pumping performance at low pressure. Those qualifications are needed for efficient embedding of the pump in the International Thermonuclear Experimental Fusion Reactor (ITER) project. As a nuclear facility (Installation Nucléaire de Base, INB) ITER is operated in accordance with the French nuclear safety regulations (INB guidelines) that demand qualification and integrity of the tritium bearing equipment to maintain performance and confinement of active gases at every stages via design, testing, commissioning, operation and surveillance. The experimental methodologies and findings are important to advocate the qualification of the pump for the purpose of radioactive light gas pumping as a roughing pump to back Torus high vacuum cryogenic pump.

Experiments and/or crystal plasticity finite element modeling of the mechanical properties of pristine and irradiated tungsten single crystal

Tungsten is used for vital applications in nuclear industry, particularly in experimental fusion reactors like ITER. Experimental evaluation of the irradiation-induced hardening is usually carried out by indentation tests. However, accurate prediction of irradiation hardening by indentation simulations faces two challenges: (1) compression data of single-crystal tungsten for indentation model calibration are lacking, since most available data are from tension tests, but tungsten exhibits a tension-compression asymmetry; (2) the irradiation hardening model for tungsten needs to be improved by incorporating more accurate irradiation defect strengthening mechanisms and defect density evolution laws that are quantitatively extracted from the lower scale modeling such as dislocation dynamics (DD) and molecular dynamics (MD). In this work, we present a comprehensive solution for the neutron irradiation hardening prediction of tungsten based on the experiments on pristine single-crystal tungsten and lower scale simulations. First, we conduct compression and Vickers indentation experiments to calibrate and validate the model for indentation simulations, and extend the model with the irradiation defect hardening law and the defect density evolution law informed by DD and MD simulations. Then, we use the irradiation hardening results from the literature to benchmark the model for its effectiveness, and find that the predicted hardness increases match well with the experimental results for different doses. Moreover, we discuss the drawback to the widely-used conventional hardening models, and find that these models tend to overestimate the defect strengthening coefficients.

A Short Communication on Progress and Problems of ITER Fusion Project

In recent years, it becomes clear that ITER project in France, as one of the largest experimental fusion reactors underway, is far away from achieving net energy production. In this review article, we presented a short communication this week with Robert Neil Boyd, a senior physicist who happens to have his own working design of fusion reactor in the past. We hope that this transcript of our communication with him (as per 15-17th Nov. 2021) may be found useful for younger scientists.

Effect of thermomechanical treatment on MX phase precipitation behavior in CLAM steel

Reduced Activation Ferritic/Martensitic (RAFM) steels are the preferred first wall structural materials in nuclear fusion experimental reactors at this stage, which puts forward high requirements for its creep properties. Previous references point out that the main factor for the creep resistance of RAFM steel is the massively precipitated MX phases in the matrix. Thermomechanical treatment (TMT) technology can promote the precipitation of a large number of nano-sized MX phases, and has great potential for improving the creep strength of RAFM steels. In this paper, China Low Activation Martensitic (CLAM) steel is taken as the research object, and four different thermomechanical treatment processes are designed for comparison. The purpose is to deepen our understanding of the effects of austenitization temperature, deformation temperature and strain on the precipitation of MX phases. It is found that the number density of MX phase after TMT increases by two orders of magnitude, and the average particle size decreases by about 79% when the austenitization temperature increases from 1000 to 1150 °C, while the effect of deformation temperature and strain is relatively small; after TMT, the specimen with the largest number and the smallest size of MX phase has the best creep performance, and vice versa.

Optimization Analysis of the Structural Design of NNBI Cryosorption Pumps

Cryosorption pumps create a vacuum by adsorbing gas at low temperature through porous solid adsorbents. The transmission probability of gas molecules and heat loads of cryosorption pumps are important factors affecting its performance. Herein, Molflow software based on the Monte Carlo principle is used to analyze the effects of the structural design of cryosorption pumps on transmission probability. The influence of structural design on radiation heat transfer is analyzed by ANSYS Steady-State Thermal software. This provides a reference for the design of a cryosorption pump to validate the prototype of a neutral beam injector for the China Engineering Fusion Experimental Reactor (CFETR).

RAMI analysis of the Collective Thomson Scattering system front-end – Part2 – reliability block diagram analysis

ITER is an experimental tokamak nuclear fusion reactor with the goal of achieving afusion energy gain factor of 10. Nuclear fusion presents an extreme environment for materials during operations while requiring continuous uninterrupted operation.The Reliability Block Diagram (RBD) is a graphic methodology for system reliability modelling which is constructed from the reliability-wise relations between different subsystems/components and their respective reliability and maintenance data. In this paper, we construct andanalyse the RBD for the Collective Thomson Scattering (CTS) front-end both in terms of fulfilling its functions and its effects in the ITER tokamak operations. Our RDB analysis results were then used to propose mitigation actions which include design changes and operational procedures to deal with the identified failure modes.Our initial results indicate that the system's mean availability at the end of its lifecycle of 20 years is no greater than 82.55% which was due to low Mean Time Between Failures (MTBF) of a critical system component – theSplit Biased Waveguide (SBWG) – aswell as several thermally loaded components. After mitigation actions, which aimed at controlling the system's exposure to heat loads, the MTBFs increased and the resulting mean availability achieved values over 97%, in conformity with design specifications. Furthermore, considering the CTS failure modes with impact on ITER operations, the expected value for the mean availability is of 99.995%. Therefore, we concluded that these failure modes did not require mitigation actions.In this article, we cover the second half of the RAMI analysis of the CTS diagnostic, the RBD analysis.

Design Studies, Magnetic Calculations and Structural Assessment For the DTT Central Solenoid

The “Divertor Tokamak Test facility” (DTT) is an experimental fusion reactor being built in Frascati (IT) in the framework of the European Fusion Roadmap. The DTT Central Solenoid operates in a pulsed regime to generate the magnetic flux needed to induce the plasma current and is therefore subjected to burdensome load conditions. In DTT, this magnet will consist of a stack of six layer-wound independently energized modules, wound with Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn Cable-in-Conduit Conductors. Different layout solutions have been studied to improve the performance of the CS in terms of magnetic flux density and structural integrity. The design presented is the result of the “Design Explorer” algorithm that has been specifically developed for this purpose. This code investigates all possible configurations to find the optimum conductor geometry and winding solution by considering measurable physical quantities to maximize the magnetic flux density while maintaining mechanical stresses at an acceptable level. This manuscript includes the electromagnetic calculation related to the heterogeneous plasma scenarios of the machine, the static structural and the fatigue life assessment and the full set of in-depth analyses performed for the qualification of the main components of the CS structure.

Progress on wide-range equation of state for hydrogen and deuterium

Hydrogen is the most abundant element in nature and an important object of astrophysics and ICF research. This paper briefly presents an overview of the research progress in wide-range equation of state and especially comments assessment of the most recent shock compression experiments on Omega laser facility and the theoretical models. Based on the previous work, the wide-range equation of state of hydrogen is constructed by using the improved chemical free energy model, the first-principle numerical simulation results and the multi-parameter equation of state model, which is applicable in the temperature range of 20−108 K and the density range of 10−7−2000 g/cm3. Compared with experimental results, such as those of shock compression experiment, static high pressure isotherm experiment and sound velocity experiment, the newly-constructed wide-range equation of state for hydrogen has high confidence and provides high precision data for astrophysics, inertial confinement fusion, international thermonuclear experimental reactor and other engineering physics designs. The construction and validation method of the hydrogen wide-range equation of state can also be applied to its isotope deuterium. In comparison with current models published abroad, the deuterium wide-range equation of state constructed by this method is in better agreement with the experimental data of principal and secondary Hugoniot published in 2019. This paper also points out the temperature-density regimesthat need attention in future research.

ОШИБКИ В РАДИАЦИОННОМ НАГРЕВЕ УРАНА ПРИ РАСЧЕТАХ С ИСПОЛЬЗОВАНИЕМ БИБЛИОТЕКИ FENDL 3.1D

The work is devoted to the analysis of uranium radiation heating results under the action of neutron radiation. The revealed results discrepancies in the radiation heating modeling of the uranium 235 and 238 using FENDL 3.1d and ENDF nuclear data libraries motivate the study. At the same time, all initial data in the computational model, namely, the neutrons energy spectrum, mass of fissile material, geometry and surrounding structures composition were identical. In the course of the work, the results of energy release calculations in uranium cells were compared using various libraries. Analytical evaluation of the radiation heating results in test cells is given. The obtained data indicate erroneous values of neutron kerma factors in FENDL 3.1d nuclear data library. Fusion Evaluated Nuclear Data Library was created by the working group of the Nuclear Data Section of the IAEA in the mid-1990s. FENDL is focused on application in the design of fusion plants as well as plants for materials testing such as IFMIF. The expanded application required the addition of some materials to the library, the expansion of the energy range to higher energies and the addition of data for reactions caused by charged particles. Version 3.1d was released in January 2018 and is widely used in nuclear analysis of structural materials for the experimental fusion reactor ITER and the demonstration fusion reactor DEMO. The results of this work have been submitted to the ITER International Organization contact group for presentation to the developers of the FENDL library.

Development and characterization of advanced neutron multiplier materials

Neutron multiplier materials are essential for self-sufficient tritium production and are closing the fuel cycle of fusion reactors. Until now, a concept of pebble bed consisting of interchanging layers of beryllium and lithium ceramic pebbles was considered for the Helium-Cooled Pebble Bed (HCPB) tritium-breeding module of the first experimental fusion reactor ITER as well as for the next demonstration fusion reactor DEMO. However, this concept depends on the availability of large amounts of pure beryllium pebbles and is also limited by its material properties like for example the tritium accumulation under irradiation.The results of tritium retention and analytical microstructural studies of beryllium pebbles obtained within the framework of the HIDOBE irradiation campaign suggest that a significant fraction of generated tritium (up to 100% below 500°C) is trapped within helium bubbles. Being negligible in the ITER tritium-breeding module (TBM), the total accumulated tritium inventory imposes severe safety issues and exceeds acceptable limits for the DEMO blanket. Therefore, advanced neutron multiplier materials such as beryllides have to be well characterized for their applicability in the HCPB blanket of DEMO and beyond. The usage of an advanced material with lower volumetric swelling, lower tritium retention, increased irradiation and chemical resistance as well as with higher melting temperature allows to switch from the pebble bed concept to a solid hexagonal block-based one. For the fabrication of titanium beryllide samples both the semi-industrial fabrication route utilizing the hot extrusion of rods and the industrial approach using vacuum hot pressing of a Be-Ti powder mixture were explored. In this contribution, we discuss reasons for the transition from pure beryllium to beryllides, respective changes of the HCPB blanket design and a successful demonstration of the feasibility of a beryllide block fabrication by an industrial method.

Interface design of the blanket system for CFETR

The China Fusion Engineering Test Reactor (CFETR) aims to bridge the gaps between fusion experimental reactor ITER and demonstration fusion power reactor. Blanket, one of the key components of the CFETR, is of great importance to achieve the functions of tritium breeding, heat extraction and radiation shielding. As for the next device for fusion energy research in China, two blanket candidate options have been developing in parallel to decrease the risk of both physical and technical issues. Apart from the research and design work on the breeding blanket module itself, the interfaces definition and integration of the blanket system is vital, particularly due to the system interdependencies. The interface design of the blanket system is a challenge due to several factors, such as the thermal expansion, the electromagnetic loads, the feasibility of remote maintenance. In this paper, a preliminary interface definition and integration design of the CFETR blanket system has been presented, based on the design with R = 7.2 m and a = 2.2 m. The purpose is to complete the interface definition and integrated design of the blanket and other components or systems. The vacuum vessel has 16 sectors, and each sector has two inboard and three outboard blanket segments. The blanket segment will be installed and removed by remote maintenance tools through the upper ports. The work addresses the interfaces between the blanket and other system, such as diagnostic system and heating system. Preliminary assessment of the impact on tritium breeding ratio (TBR) for the candidate water cooled ceramic breeder blanket integration scheme has been presented highlighting the influence of the auxiliary interfaces system to TBR.

The CFETR tritium plant: Requirements and design progress

Abstract A tritium plant is essential for successful and safe operation of the China Fusion Engineering Test Reactor (CFETR) which aims to bridge the gaps between a fusion experimental reactor (namely ITER) and a demonstration reactor. At the end of 2017, CFETR integration engineering design started, which will last until the end of 2020. During the engineering design phase, the CFETR tritium plant is foreseen to accomplish a detailed conceptual design. At the end of 2014, a pre-conceptual design of tritium plant was already accomplished for the early CFETR concept which was based on a small size Tokamak (200 MW). However, the CFETR concept has changed a lot from the early version to the current one which was based on a larger Tokamak (200−1500 MW). A notable change is that the current CFETR concept will cover both the engineering test phase and the DEMO validation phase. As a result, the pre-conceptual design of CFETR tritium plant is considered to be out of date. In this report, the interfaces and requirements of CFETR tritium plant are redefined, and the most recent design progress of CFETR tritium plant is presented.

Effect of magnetic field environment on the performance of 3/2 solenoid valve

• 3/2 solenoid valve functionality has been checked under external magnetic field environment. • External magnetic field limit has been defined below which 3/2 solenoid valve functions well. • Magnetic shielding of the 3/2 solenoid valve has been proposed. ITER, an experimental fusion reactor uses solenoid valves to close cryogenic control valves instantly in case of any incidental event in the ITER cryogenic system takes place. Many of these solenoid valves are functioning with the magnetic field environment of the order of ∼25 m T. Since a solenoid valves is an electrically activated valve, presence of the environment magnetic field can lead to partial or non-closing of the valve. In this paper, the effect of magnetic field environment on the performance of 3/2 solenoid valves through simulation and experiment is studied. The variable external magnetic field is generated by supplying variable power from Switched Mode Power Supply (SMPS) to the magnetic coil and magnetic field readings are recorded with a tesla meter. The validation of the magnetic field profile of the magnetic coil is done by comparing the E-mag analysis simulation performed on ANSYS® platform and the experimental results. An experiment setup, inside ITER-India cryogenic laboratory, has been assembled to see the functionality of the 3/2 solenoid valve in the presence of an external magnetic field. A 3/2 solenoid valve is placed in the external magnetic field which has been generated with the help of an air core inductor magnetic coil. The external magnetic field from the coil is varied from 10 m T to 35 m T with the help of a variable power supply and checked the functionality of the solenoid valve in different orientation and different external magnetic fields strengths. The outcome of the study shows that the 3/2 solenoid valve stopped functioning over the external magnetic field strength of 14.29 m T. The paper also concludes the requirement of the magnetic shielding and correct orientation of the solenoid valves to be installed in order to nullify the effect of external magnetic field on the functionality of the 3/2 solenoid valves.

Conceptual Design of 15 Tesla Conductor Test Facility for Future Fusion Reactor

“China Fusion Engineering Test Reactor (CFETR)” is a new generation of tokamak which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). The superconducting magnet system is the core component in tokamak and future fusion reactor, and the superconducting magnet system for CFETR will be a big challenge due to the higher magnetic field is required. Based on the current design, the maximum field of Toroidal Field (TF) coil is around 14.5 T, and for Center Solenoid (CS) coil, the maximum field would be even higher. Now the significant work is developing high-performance Cable-in-Conduit Conductor (CICC) for these high field coils. Obviously, a test facility is necessary for developing new CICCs. It is well known that SULTAN facility plays an important and successful role in ITER conductor development, almost all ITER sample conductors were tested at SULTAN. But it's hard to satisfy future test work due to its magnetic field limitation (<12 T). Now in China, a new program has been launched to build a new conductor test facility which has similar functions to SULTAN, but can provide background field up to 15 T. This paper will introduce the concept design of the magnet system of the superconducting conductor test facility.

Beam energy recovery for fusion and collector design for tests on compact sources

The next fusion project DEMO, which will be the evolution of the experimental fusion reactor [International Tokamak Experimental Reactor (ITER)], would require a high efficient energy production. As in ITER, DEMO will use fast Neutral Beam (NB) injectors to increase the plasma temperature needed for the fusion reaction. A way to recover the electric energy production efficiency in DEMO could be the beam energy recovery in the NB production, which is produced by a D- beam, neutralized by a gas cell with 60% efficiency. A compact energy recovery device with an axisymmetric cylindrical ion collector that uses only decelerating electric fields combined with the beam space charge effect has been recently proposed. It can be used for a test on the beam of the NIO1 (Negative Ion Optimization 1) source, a compact ion source (scaled down from ITER size sources) that has been developed at INFN-LNL and Consorzio RFX (Padua). The detailed collector design to be used on one of the beamlets of the NIO1 source within typical space limitation is presented and discussed here. Furthermore, a preliminary trajectory simulation for a beam with a rectangular geometry similar to the beam used in ITER to verify the beam recovery for a nonaxial symmetric geometry is also shown.

A flexible and compact electronic system for the bolometric diagnostic with on-line calibration capability

At ENEA Frascati laboratories an updated electronic system for the bolometric diagnostic is under development. The system is based on the most recent electronic technology and it is able to manage a large number of channels. It is flexible, reliable and compact in order to be suitable for applications on large fusion experimental machines or reactors. The system has been conceived to be applied to bolometric detectors based on a resistive bridge, it is mainly based on the digital signal processing technology to guarantee high accuracy, large dynamic range and a high signal to noise ratio. It is also able to manage a large offset of the bridges. A further important feature of the system is the embedded calibration capability of the sensor detectors that allows for the calibration of each channel in parallel and in a very short time with the bolometer heads installed and without any manual intervention on the lay-out, thus also allowing for calibration just shortly before the plasma experiment. A further fast embedded on-line calibration capability allows for the estimation of the sensor temperature and of the temperature gradient in the bolometer head, inside the bolometer measurement temporal window. The paper describes the main concepts, the general architecture of the system and in detail the calibration and the on-line calibration functionalities.

Conceptual design of Japan’s fusion DEMO reactor (JADEMO) and superconducting coil issues

Goals of Japan’s fusion demonstration (DEMO) reactor are to demonstrate (1) steady and stable electric power generation in a power plant scale, (2) self-sufficient production of fuel (tritium), and (3) reasonable availability using a remote maintenance scheme anticipated in a commercial plant. Main design parameters of JA DEMO are a plasma major radius of 8.5 m, fusion output of 1.5-2 GW, magnetic field on the plasma axis of 5.94 T. The superconducting coil system of the reactor consists of a central solenoid (CS), 7 poloidal field (PF) coils and 16 toroidal field (TF) coils. Regarding CS and PF coils, superconducting coil technology on DEMO is basically the same as that on the world largest fusion experimental reactor called ITER. In contrast, TF coils have a technology gap on magnetic energy and the resulting stress between ITER and DEMO due to their size and magnetic field. In particular, the necessity of higher design stress is critical for TF coils, requiring the development of high strength cryogenic steels surpassing the existing ones. The fundamental design strategy to mitigate tolerances in TF coil fabrication is also presented.

In Vitro Analysis of the Effects of ITER-Like Tungsten Nanoparticles: Cytotoxicity and Epigenotoxicity in BEAS-2B Cells.

Tungsten was chosen as a wall component to interact with the plasma generated by the International Thermonuclear Experimental fusion Reactor (ITER). Nevertheless, during plasma operation tritiated tungsten nanoparticles (W-NPs) will be formed and potentially released into the environment following a Loss-Of-Vacuum-Accident, causing occupational or accidental exposure. We therefore investigated, in the bronchial human-derived BEAS-2B cell line, the cytotoxic and epigenotoxic effects of two types of ITER-like W-NPs (plasma sputtering or laser ablation), in their pristine, hydrogenated, and tritiated forms. Long exposures (24 h) induced significant cytotoxicity, especially for the hydrogenated ones. Plasma W-NPs impaired cytostasis more severely than the laser ones and both types and forms of W-NPs induced significant micronuclei formation, as shown by cytokinesis-block micronucleus assay. Single DNA strand breaks, potentially triggered by oxidative stress, occurred upon exposure to W-NPs and independently of their form, as observed by alkaline comet assay. After 24 h it was shown that more than 50% of W was dissolved via oxidative dissolution. Overall, our results indicate that W-NPs can affect the in vitro viability of BEAS-2B cells and induce epigenotoxic alterations. We could not observe significant differences between plasma and laser W-NPs so their toxicity might not be triggered by the synthesis method.