Fusion Nuclear Environment Research Articles

Inverse kinematics for the CMOR in the CFETR based on workspace analysis: A novel approach

This paper delves into the intricate field of inverse kinematics in the context of the Chinese Fusion Engineering Test Reactor (CFETR) and its specialized robot, the multi-purpose overload robot (CMOR). Nuclear fusion, a cornerstone for sustainable energy, presents immense potential with benefits like high energy output and minimal environmental impact. The goal of the CFETR is to transform nuclear fusion energy from theory to practical solutions to solve the global energy crisis. The CMOR, with nine degrees of freedom, is pivotal in maintaining in-vessel components damaged by intense radioactive exposure. This study introduces a novel approach to solve the inverse kinematics of the CMOR based on the CFETR's work environment. It involves dividing the working space, identifying target interval based on coordinates, and employing intelligent algorithms for precise and efficient kinematic resolution. Six algorithms were evaluated, highlighting the proposed approach's superior computational speed and accuracy, crucial for real-time monitoring and control in CFETR operations. This work presents a significant advancement in redundant robotic kinematics in nuclear fusion environment.

ITER Test Blanket Module—ALARA Investigations for Port Cell Pipe Forest Replacement

The objective of the ITER test blanket module (TBM) program is to provide experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. The ITER test blanket modules are installed and operated inside the vacuum vessel (VV) at the equatorial ports located within port plugs (PP), and each PP includes two TBMs. After each 18-month-long plasma operation campaign, the TBM research plan testing program requires the replacement of the TBMs with new ones during the ITER long-term shutdown, called long-term maintenance (LTM). The replacement of a TBM requires the removal/reinstallation of all test blanket system (TBS) equipment present in the port cell (PC), including those in the port interspace (PI), called pipe forest (PF). TBSs shall be designed so that occupational radiation exposure (ORE) can be as low as reasonably achievable (ALARA) over the life of the plant to follow the ITER policy. To implement ALARA process requirements, design activities shall consider careful integration investigations starting from the early phase to address all engineering aspects of the replacement sequence. The case study focuses on the PF replacement, in particular the port cell operations. This paper describes the investigations and findings of the ALARA optimisation process implementation in the early engineering phase of the PF.

Microstructural evolution in pure and ZrC strengthened tungsten under ion irradiation at 600°C

The microstructural evolution in ultra-fine grain (UFG) tungsten reinforced with ZrC particles under ion irradiation was investigated by transmission electron microscopy and compared to the effect of ion irradiation in reference ITER specification tungsten. Here, we report the evolution of the microstructure under heavy ion irradiation to mimic the neutron irradiation damage taking place in a nuclear fusion environment. The irradiation was performed with 9.1 MeV gold ions at 600°C to reach the end-of-life irradiation dose in ITER and DEMO divertor (i.e. 1 and 5 dpa). The microstructure is investigated by means of transmission electron microscopy. It is found that the ion induced defect structure does not deviate significantly from the reference tungsten. The main defects are a/2<111> type dislocation loops and nano-metric voids. The defect concentration is already saturated at 1 dpa and the higher irradiation damage was not found to have a significant effect on the defect concentration.

Plastic deformation in advanced tungsten-based alloys for fusion applications studied by mechanical testing and TEM

In this work, we have assessed mechanical properties of several tungsten grades considered as perspective materials for applications in plasma facing components in the nuclear fusion environment, where the neutron irradiation damage is expected to cause embrittlement. In particular, the work focuses on two aspects: bending tests to deduce the onset of ductile deformation and microstructural analysis of the reference and plastically deformed materials. The microstructure in the reference state and the one induced by plastic deformation at 600 °C is studied by means of transmission electron microscopy (TEM). Six different types of tungsten-based materials were assessed: two commercial grades produced according to ITER specifications in Europe and China and four lab-scale grades utilizing different reinforcement options. The comparative assessment of tensile and bending strength was performed at 600 °C accompanied with a detailed TEM analysis. The deformation-induced microstructure was characterized and compared for all studied grades in terms of the dislocation density, heterogeneity, observation of pile-ups and tangles specifically near grain boundaries and/or strengthening particles. The obtained data will serve as reference information required to assess the impact of neutron irradiation.

3D modelling of MHD mixed convection flow in a vertical duct with transverse magnetic field and volumetric or surface heating

Understanding phenomena associated with the multiple effects/interactions of the fusion nuclear environment on liquid metal flow is required to correctly design liquid metal (LM) blankets for fusion facilities. These effects are investigated in the present work by numerically simulating 3D LM MHD flow. The simulated geometry consists of a straight, vertical duct which runs perpendicular to a strong, fringing applied magnetic field. There is also a region of applied heating as the primary goal is to explore buoyancy effects in MHD duct flows. Results are presented for both buoyancy assisted (upwards) and buoyancy opposed (downwards) flows in conducting and insulating ducts for a range of Hartmann numbers (Ha) up to 100, Reynolds numbers (Re) from 103 to 104 and Grashof (Gr) numbers from 107 to 108. While increasing Gr or decreasing Re increases buoyancy effects, increasing Ha was shown to increase maximum temperature through turbulence reduction. The extent to which the MHD mixed convection flows are quasi-2D is analyzed and buoyant effects, in competition with electromagnetic forces, are shown to bring about 3D flow features not seen in purely MHD flows. Volumetric nuclear heating with steep gradients is applied to the vertical MHD flows for comparison to flows with surface heating only. Surface heating generates stronger buoyancy effects than volumetric heating of the same total power; however, many of the same phenomena occur. Therefore, surface heating, the only option for lab experiments, can provide indication of the effects of volumetric heating in MHD flows.

Impact of neutron irradiation on the strength and ductility of pure and ZrC reinforced tungsten grades

The strength and ductility of two tungsten products, developed for application in nuclear fusion environment, are studied before and after neutron irradiation using uniaxial tensile tests. The first product is a commercially pure tungsten, produced by AT&M company according to ITER specification, and the second one is reinforced with zirconium carbide (W–ZrC) particles. The addition of ZrC particles leads to a reduction of the ductile to brittle transition temperature (DBTT) in non-irradiated conditions down to 50–100 °C without loss of strength and of other attractive properties of tungsten.The neutron irradiation was performed in the range 625–700 °C up to 1.125 dpa. The tests were performed to screen the shift of the DBTT as well as to characterize the evolution of the strength and ductility at the irradiation temperature. In addition, a series of interrupted tensile tests were performed in order to determine the variation of the yield strength as a function of temperature using an original single specimen test method.The neutron irradiation causes the reduction of the total elongation of both tungsten products. The DBTT range, which was evaluated from the tensile test results, of W–ZrC lies in the 300–500 °C range (while it is ∼100 °C in non-irradiated state). The DBTT range of pure tungsten is between 400 and 575 °C i.e. higher than that of W–ZrC. The irradiation hardening, measured at ∼600 °C, which is close to the irradiation temperature, leads to an increase of the proof stress by a factor of two in both studied grades. Despite that the irradiation induced hardening, both products retain a total elongation of about 10% prior to fracture. W–ZrC exhibits a similar total elongation at 500 °C, thus maintaining a significant ductility resource, while pure W becomes brittle at 500 °C and below.

Towards V-based high-entropy alloys for nuclear fusion applications

By mixing elements with favourable nuclear activation properties to create high-entropy alloys, it may be possible to create a material that can withstand a nuclear fusion environment while minimising the radioactive waste produced. Such a material could be used in the extreme thermal and irradiation conditions of a fusion blanket. A suite of previously unexplored V–Cr–Mn and V–Cr–Mn–Ti alloys have been fabricated then homogenised and the resultant microstructures and phases were characterised. Results demonstrate that single-phase body centred cubic solid solution microstructures can be formed in highly-concentrated alloys incorporating low-activation elements, which is promising for a fusion alloy development standpoint.

Outer Divertor Damage Characterization from Deuterium Plasma Bombardment in Graphene-Coated Tungsten in the C-2W Device

This research explores the performance of graphene as a coating for plasma-facing components (PFCs) in a nuclear fusion environment. Our recent studies have shown that graphene can act as a resistant layer against plasma exposure and ion bombardment. PFCs tend to develop surface morphologies that lead to mass loss of the wall material, potentially diminishing their lifetime and degrading plasma performance. We present a characterization of graphene-coated samples of W irradiated in the C-2W divertor. Energy analyzers were used to determine average ion fluxes to the samples on the order of 1018 D+/cm2. Two samples were exposed over 1210 plasma discharges. Raman spectroscopy showed that slow ions (30 < E < 100 eV) interact strongly with the graphene, introducing vacancies into the membrane (ID/IG ~ 0.7), making it possible to assess the limiting factors on such a coating’s lifetime. We also found that graphene slows down impurity deposition on the material surfaces due to graphene’s stable configuration and low surface energy. This first attempt at testing the coating in a large-scale fusion experiment aims to expand the possible wall candidates for PFCs.

Thermal and Electromagnetic Transients in Liquid Metal Surfaces of the FNSF

The Fusion Nuclear Science Facility (FNSF) is an intermediate step in the path to commercial fusion energy that will accommodate the extreme fusion nuclear environment and the complex integration of components and their environment as well as the relevant nuclear science and plasma physics. The transient thermal and electromagnetic loads on plasma-facing components in FNSF have been shown to offer significant design challenges that are difficult to meet with solid walls. Hence, the project team is investigating the feasibility of using liquid walls to ameliorate some of the risk associated with solid wall designs.In this paper, we examine the effects these transient loads will have on a liquid wall. Mass loss is considered using standard evaporation models accounting for transient surface temperatures. The heat transfer is modeled with a one-dimensional transient conduction model that accounts for evaporative losses. No liquid motion is considered. Loss rates of tens of microns per edge-localized mode (ELM) are predicted. Peak heat fluxes are treated parametrically to help address the substantial uncertainty inherent in models for the timing and spatial distribution of the heat deposited during the ELM. Boiling is considered but is found to not be of consequence, as the temperatures required for homogeneous nucleation of bubbles are substantially higher than a conventional boiling point. It should be noted that all evaporation calculations are for evaporation into a vacuum. In the future, we intend to incorporate these evaporation rates into an edge physics code to self-consistently model the net mass flows at the liquid surface in a tokamak.Electromagnetic effects due to ELMs and disruptions are accounted for by assuming a stationary plasma quench. ELMs are addressed assuming a small fluctuation in the plasma current during an event, while disruptions are addressed assuming a full quench of the current. The variation in the plasma current induces currents in the conducting fluid, leading to forces on the liquid (and subsequent motion). A commercial finite element code is used to calculate the induced currents and forces associated with a static liquid divertor. Liquid motion is not considered in this calculation, so no magnetohydrodynamic (MHD) currents are addressed, but a simplified model is presented to estimate the impact of these currents on the liquid motion. Based on these calculations, the acceleration of the liquid is expected to be quite high, and containment of the liquid is likely not possible. The MHD effects appear to be relatively minor.

Overview of the fusion nuclear science facility, a credible break-in step on the path to fusion energy

The Fusion Nuclear Science Facility (FNSF) is examined here as part of a two step program from ITER to commercial power plants. This first step is considered mandatory to establish the materials and component database in the real fusion in-service environment before proceeding to larger electricity producing facilities. The FNSF can be shown to make tremendous advances beyond ITER, toward a power plant, particularly in plasma duration and fusion nuclear environment. A moderate FNSF is studied in detail, which does not generate net electricity, but does reach the power plant blanket operating temperatures. The full poloidal Dual Coolant Lead Lithium (DCLL) blanket is chosen, with alternates being the Helium Cooled Lead Lithium (HCLL) and Helium Cooled Ceramic Breeder/Pebble Bed (HCCB/PB). Several power plant relevant choices are made in order to follow the philosophy of targeted technologies. Any fusion core component must be qualified by fusion relevant neutron testing and highly integrated non-nuclear testing before it can be installed on the FNSF in order to avoid the high probability of constant failures in a plasma-vacuum system. A range of missions for the FNSF, or any fusion nuclear facility on the path toward fusion power plants, are established and characterized by several metrics. A conservative physics strategy is pursued to accommodate the transition to ultra-long plasma pulses, and parameters are chosen to represent the power plant regime to the extent possible. An operating space is identified, and from this, one point is chosen for further detailed analysis, with R=4.8m, a=1.2m, IP=7.9 MA, BT=7.5T, βN<2.7, n/nGr=0.9, fBS=0.52, q95=6.0, H98 ∼1.0, and Q=4.0. The operating space is shown to be robust to parameter variations. A program is established for the FNSF to show how the missions for the facility are met, with a He/H, a DD and 5 DT phases. The facility requires ∼25years to complete its DT operation, including 7.8 years of neutron production, and the remaining spent on inspections and maintenance. The DD phase is critical to establish the ultra-long plasma pulse lengths. The blanket testing strategy is examined, and shows that many sectors have penetrations for heating and current drive (H/CD), diagnostics, or Test Blanket Modules (TBMs). The hot cell is a critical facility element in order for the FNSF to perform its function of developing the in-service material and component database. The pre-FNSF R&D is laid out in terms of priority topics, with the FNSF phases driving the time-lines for R&D completion. A series of detailed technical assessments of the FNSF operating point are reported in this issue, showing the credibility of such a step, and more detailed emphasis on R&D items to pursue. These include nuclear analysis, thermo-mechanics and thermal-hydraulics, liquid metal thermal hydraulics, transient thermo-mechanics, tritium analysis, maintenance assessment, magnet specification and analysis, materials assessments, core and scrape-off layer (SOL)/divertor plasma examinations.

Overview of ITER TBM program objectives and management

One of the most important ITER missions is to test tritium breeding blanket (TBB) concepts that would result in tritium self-sufficiency, an extraction of high grade heat and net electricity production in future fusion reactors. Knowledge of the TBB performance characteristics in the integrated fusion nuclear environment is essential to designing the TBB of the DEMO and the fusion reactor. Three equatorial ports in ITER are allotted for testing test blanket modules (TBMs), which are mock-ups of the TBB, allowing up to six TBMs to be simultaneously installed and tested – two TBMs per each port. Conceptual designs for all TBM systems are now completed, and the important next milestones are the achievement of the preliminary and the final design by 2025. The TBM systems will be installed during Assembly Phase III, which is planned to start in 2030, and tested during the ITER Pre-Fusion Power Operation 2, which is planned to start in 2032. This paper describes the objectives and management aspects of the TBM Program. Copyright © 2017 John Wiley & Sons, Ltd.

Nuclear Analyses of Indian LLCB Test Blanket System in ITER

Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

Design and characterization of a prototype divertor viewing infrared video bolometer for NSTX-U.

The InfraRed Video Bolometer (IRVB) is a powerful tool to measure radiated power in magnetically confined plasmas due to its ability to obtain 2D images of plasma emission using a technique that is compatible with the fusion nuclear environment. A prototype IRVB has been developed and installed on NSTX-U to view the lower divertor. The IRVB is a pinhole camera which images radiation from the plasma onto a 2.5 μm thick, 9 × 7 cm2 Pt foil and monitors the resulting spatio-temporal temperature evolution using an IR camera. The power flux incident on the foil is calculated by solving the 2D+time heat diffusion equation, using the foil's calibrated thermal properties. An optimized, high frame rate IRVB, is quantitatively compared to results from a resistive bolometer on the bench using a modulated 405 nm laser beam with variable power density and square wave modulation from 0.2 Hz to 250 Hz. The design of the NSTX-U system and benchtop characterization are presented where signal-to-noise ratios are assessed using three different IR cameras: FLIR A655sc, FLIR A6751sc, and SBF-161. The sensitivity of the IRVB equipped with the SBF-161 camera is found to be high enough to measure radiation features in the NSTX-U lower divertor as estimated using SOLPS modeling. The optimized IRVB has a frame rate up to 50 Hz, high enough to distinguish radiation during edge-localized-modes (ELMs) from that between ELMs.

TPE upgrade for enhancing operational safety and improving in-vessel tritium inventory assessment in fusion nuclear environment

The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γimax of >1.0×1023m−2s−1 and qheat of >1MWm−2 with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

Modeling and simulation of helium behavior in tungsten: A first-principles investigation

Tungsten (W) is regarded as the most promising candidates for plasma-facing materials (PFMs) in future fusion reactors. However, W will be exposed to high fluxes of helium (He) plasma in a nuclear fusion environment, which will degrade the thermal and mechanical properties of W. We have investigated the behavior of He in W using a first-principles method based on density functional theory. It has demonstrated that the quantitative stress indicator and the qualitative hard-sphere model can be used to characterize the effective volume and the stability of interstitial He in W. The hydrostatic tensile (compressive) strain will enhance (suppress) the dissolution and diffusion of He in W. The existing of He will have negative effect on the ideal strength of W. Vacancies and grain boundaries defects in W can serve as trapping centers for He, because the defects can provide large space for He. He will spontaneously form He clusters at the near surface by self-trapping. The pre-existing impurity/alloying element can significantly reduce the solution energy of He in W due to the electron density redistribution induced by them, such as C and Nb. The results will provide good reference for developing W materials as PFMs. This paper reviews our recent findings regarding the behavior of He in W.

Thermodynamic and kinetic properties of intrinsic defects and Mg transmutants in 3C–SiC determined by density functional theory

Density functional theory (DFT) is used to calculate the thermodynamic and kinetic properties of transmutant Mg in 3C–SiC due to high-energy neutron irradiation associated with the fusion nuclear environment. The formation and binding energies of intrinsic defects, Mg-related defects, and clusters in 3C–SiC are systematically calculated. The minimum energy paths and activation energies during point defect migration and small cluster evolution are studied using a generalized solid-state nudged elastic band (G-SSNEB) method with DFT energy calculations. Stable defect structures and possible defect migration mechanisms are identified. The evolution of binding energies during Mg2Si formation demonstrates that the formation of Mg2Si needs to overcome a critical nucleus size and nucleation barrier. It is found that C vacancies promote the formation of the Mg2Si nucleus, and formation of which results in a compressive stress field around the nucleus. These data are important inputs in meso- and macro-scale modeling and experiments to understand and predict the impact of Mg on phase stability, microstructure evolution, and performance of SiC and SiC-based materials during long-term neutron exposures.

Transmutation of silicon carbide in fusion nuclear environment

The amount and type of metallic transmutants produced in SiC/SiC when used in magnetic (MFE) and inertial (IFE) confinement fusion systems are determined and compared to those obtained following irradiation in fission reactors. Up to ∼1.3% metallic transmutants are generated at the expected lifetime of the fusion blanket. Irradiation in fission reactors to the same fast neutron fluence produces about an order of magnitude lower metallic transmutation products than in fusion systems. While the dominant component in fusion systems is Mg, P is the main transmutation product in fission reactors. The impact on the SiC/SiC properties is not fully understood. The results of this work will help guide irradiation experiments in fission reactors to properly simulate the conditions in fusion systems by possible ion implantation. In addition, the results represent a necessary input for modeling activities aimed at understanding the expected effects on properties.

Structural Analysis of an Optimally Designed Spherical Tokamak Centerpost

The realization of commercialized fusion power will involve the development of new materials that can withstand the uniquely harsh nuclear fusion environment. Of particular interest are those materials that are closest to the plasma. The combination of thermal loading, neutron damage, material sputtering, and redeposition provides uniquely hostile conditions under which no material testing has yet occurred. An experimental Fusion Nuclear Science Facility is required that will create the environment that simultaneously achieves high-energy neutrons and high ion fluence necessary in order to bridge the gaps from ITER to the realization of a fusion nuclear power plant. One concept for achieving this is a high-duty-cycle spherical tokamak (ST). The centerpost is a critical component of the ST design, as it controls the size of the entire reactor. The centerpost will experience significant thermal loading and thermal gradients from ohmic heating, nuclear heating, and water cooling. Nuclear heating will also produce embrittlement and swelling in the centerpost. In addition to thermal loads, the centerpost must be designed to carry mechanical loads produced from the various magnetic fields (TF, PF, and plasma currents), both steady state and transient. The centerpost temperature must remain low enough to permit water cooling, and stresses must remain low enough so that the centerpost remains structurally sound. This study will focus on the stress analysis of a centerpost optimized to reduce the thermal gradients in the cross section.

Overview of the ITER TBM Program

The objective of the ITER TBM Program is to provide the first experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. Such information is essential to design and predict the performance of DEMO and future fusion reactors. It foresees to test six mock-ups of breeding blankets, called Test Blanket Module (TBM), in three dedicated ITER equatorial ports from the beginning of the ITER operation. The TBM and its associated ancillary systems, including cooling system and tritium extraction system, forms the Test Blanket System (TBS) that will be fully integrated in the ITER machine and buildings. This paper describes the main features of the six TBSs that are presently planned for installation and operation in ITER, the main interfaces with other ITER systems and the main aspects of the TBM Program management.

Fusion Nuclear Science Facility (FNSF) before Upgrade to Component Test Facility (CTF)

The compact (R0~1.2-1.3m) Fusion Nuclear Science Facility (FNSF) is aimed at providing a fully integrated, continuously driven fusion nuclear environment of copious fusion neutrons. This facility would be used to test, discover, and understand the complex challenges of fusion plasma material interactions, nuclear material interactions, tritium fuel management, and power extraction. Such a facility properly designed would provide, initially at the JET-level plasma pressure (~30%T2) and conditions (e.g., Hot-Ion H-Mode, Q<1)), an outboard fusion neutron flux of 0.25 MW/m2 while requiring a fusion power of ~19 MW. If and when this research is successful, its performance can be extended to 1 MW/m2 and ~76 MW by reaching for twice the JET plasma pressure and Q. High-safety factor q and moderate-plasmas are used to minimize or eliminate plasma-induced disruptions, to deliver reliably a neutron fluence of 1 MW-yr/m2 and a duty factor of 10% presently anticipated for the FNS research. Success of this research will depend on achieving time-efficient installation and replacement of all internal components using remote handling (RH). This in turn requires modular designs for the internal components, including the single-turn toroidal field coil center-post. These device goals would further dictate placement of support structures and vacuum weld seals behind the internal and shielding components. If these goals could be achieved, the FNSF would further provide a ready upgrade path to the Component Test Facility (CTF), which would aim to test, for ≤6 MW-yr/m2 and 30% duty cycle, the demanding fusion nuclear engineering and technologies for DEMO. This FNSF-CTF would thereby complement the ITER Program, and support and help mitigate the risks of an aggressive world fusion DEMO R&D Program. The key physics and technology research needed in the next decade to manage the potential risks of this FNSF are identified.