Fusion Facilities Research Articles

A Neutron Activation Method of Deuterium–Tritium Plasma Diagnostics

A modification of the neutron activation method for the diagnostics of deuterium–tritium plasma of powerful pulsed fusion facilities is proposed. The method allows one to determine the effective plasma ion temperature by width of the spectral distribution of generated DT neutrons. The physical basis of the method and conditions for its implementation are described and methods for the calibration of measuring channels sensitivity are considered.

Study the Flux and Energy of the Helium Ions Beam Produced by Two Different Dense Plasma Focus Devices When Gas Pressure Changes

In this study, the flux and energy of the helium ions beam produced by United Nations University/International Centre for Theoretical Physics Plasma Fusion Facility (UNU/ICTP PFF) and NX2 dense plasma focus devices, were studied when the pressure of gas changed by carrying out numerical experiments using Lee code. The results showed that the value of the ions beam flux increased with increasing the gas pressure until reaching a maximum value 1.463E+28 ions. m-2s-1 at 7 Torr in UNU/ ICTP PFF device while in NX2 device the maximum value is 6.442E+27 ions. m-2s-1 at 14 Torr and then decreasing while continuing pressure increasing, but the energy of ions beam showed the opposite behavior where the maximum value of energy of the helium ions beam was 13 Joul at 5 Torr in UNU/ICTP PFF device while it was 144 Joul at 14 Torr in NX2 device this is due to the effect of the geometric dimensions on the value of the induced voltage within the plasma pinch.

Conceptual Design of Irradiation Facility with 6 MeV and 7 MeV Gamma Rays at the JSI TRIGA Mark II Research Reactor

Activated cooling water in nuclear facilities can present a significant radiation source around primary cooling system causing radiation damage to electrical components, increasing doses to personnel and in the case of fusion facilities additional heating to superconducting coils. As there are only few sources of gamma rays with energies in the range of 6 MeV and 7 MeV an irradiation system using activated cooling water as the source of energetic gamma rays is proposed at the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor. Two different conceptual designs, one utilizing central irradiation channel and one utilizing radial piercing port for water activation, are presented and analysed in the paper. Despite an order of magnitude higher water activation in central channel compared to radial piercing port the 16N decay rate in the irradiation facility is comparable between both design (order of 108 decays per second) due to longer transient time from central channel to irradiation facility. In the irradiation facility the expected biological dose rates due to the 16N decay rate are in order of several mSv/h. From the results he conceptual design utilizing the radial piercing port currently presents the best option for the irradiation facility due to the simpler design of the irradiation loop, already present shielding of the loop and comparable number of 16N decay rates to central channel.

THE LINEAR ACCELERATORS REVIEW

The stationary linear plasma simulators was reviewed, which are used to study the interaction between plasma and surface with regard to fusion facilities and modeling of their edge plasma. The interaction between plasma and surface is an important issue in creation of protective materials for the International Thermonuclear Experimental Reactor (ITER). The first wall and divertor are the key plasma faced components. The basic functions of the divertor are a heat flux absorption released from the edge plasma and helium removal, which is a combustion product in the thermonuclear synthesis reaction. Relatively small facilities with limited capacities as well as large linear simulators with a capacity of several dozen of kW and various diagnostic tools have been considered separately. Since the modern linear installations have a temperature and density of the generated plasma close to SOL plasma (scrap-off layer), it is possible to research the processes which directly affect materials of the components. This is a modification of surface during plasma irradiation or with charged particle beams. This includes atomization of a surface and studying of erosion, penetration of foreign particles into the surface or coating, surface relief modification, blister generation.

Creation of Thermonuclear Experimental Reactor ITER as an Example of International Scientific and Technical Cooperation in Energy Sector

The article analyses the prospects for the implementation of the largest international scientific and technical project (“mega-science” project) ITER of controlled thermonuclear fusion. With the depletion of fossil fuel reserves, the role of new energy sources, including those that have not yet been developed, is increasing. Due to the extremely high costs and ambitious scientific goals facing the project, its practical implementation and funding are possible in close international scientific and technical cooperation in the energy sector. The role of an international group of seven participants (Russia, the EU, Japan, China, India, the Republic of Korea and the USA) in the creation of laboratory fusion facilities is noted. The choice of a Tokamak-type design for the ITER international reactor being built in France is the recognition of Russian’s decisive role. The article reveals the main parameters of international partnership, the specification of which is carried out on a regular basis at ITER coordination meetings with the participation of representatives of national ITER agencies. The parameters of interaction with scientific institutions of countries that are not members of the consortium are given. The evolution of the project since its launch in 1985 as a joint Soviet-American initiative is shown. Four options for the initial placement of the reactor are considered: in Spain, France, Canada and Japan. The features of the organizational and legal regulation of the international consortium for the management of ITER are disclosed, including the establishment of the ITER International Organization for Thermonuclear Energy for the joint implementation of the ITER project in 2006. The role of the in-kind contribution of countries in international projects in the exchange of unique world-class technologies is discussed. The main goals set by the countries participating in the project and the time horizons for their practical achievement are listed. The parameters of Russia's participation in the project are given, the prospects for further cooperation in this area are evaluated. Particular attention is paid to the peculiarities of the political interaction of the participating countries, as well as to the intermediate results already achieved during the implementation of the ITER project.

Hydrogen Retention in Tungsten Alloys Developed for Fusion Facilities (Review)

This review is focused on the effect of doping with various elements on retention in tungsten. Although tungsten is one of the most promising plasma-facing materials (PFMs), it still has a number of drawbacks. Doping is proposed as a way to make tungsten more suitable for PFM applications. A considerable number of tungsten alloys have already been developed in the course of fusion research. Doping may alter the parameters of hydrogen retention in tungsten, and this should be taken into account in selecting the materials for fusion facilities.

Mechanical Properties and Material Code of Cryogenic Structural Materials for Toroidal Field Coils used in Fusion Facilities

Mechanical Properties and Material Code of Cryogenic Structural Materials for Toroidal Field Coils used in Fusion Facilities

D1SUNED system for the determination of decay photon related quantities

The neutron fields alter the radioactive inventory of the irradiated materials leading to subsequent decay photon fields. In some cases, these fields are of relevance either intended or undesired, normally involving safety and economics aspects. The determination of these fields can be of paramount complexity if high spatial resolution is required. The determination of these fields requires both radiation transport and activation calculations. The Direct-one-Step methodology, under the assumption that the radioactive inventory activity is lineal with the neutron flux, can address the problem with only one coupled neutron-photon transport calculation. In this paper the D1SUNED code for the calculation of decay photon field and related quantities using D1S methodology is presented. Calculation capabilities including the determination of 3D decay photon sources, filtering options, and other relevant features are presented. In terms of computational load, D1SUNED, which is based on MCNP5 code, presents improvements with respect to MCNP. It can save a 79% of the RAM memory used to store the geometry, a 98% of the loading time, and an acceleration of a factor two by controlling the decay photon emission, boosting the simulations for ITER-like problems. D1SUNED has been validated with the FNG benchmark experiment considering the null hypothesis rejection test and the C/E ratio with very positive results. As a consequence, D1SUNED has become a reference tool for the design of ITER, and other relevant nuclear fusion facilities.

Capabilities of GaN/AlN/GaN structures as high-intensity pyroelectric laser sensors

The use of transparent Al2O3/GaN/AlN/GaN structures as pyrometric sensors for measuring the parameters of high-intensity laser pulses is proposed. The peculiarities of the employment of such sensors in laser fusion facilities are analysed. Post-pulse distributions of the absorbed energy density are obtained for various parameters of both GaN layers. The local maxima of these distributions are minimised by varying the ratio of donor concentration and the ratio of their thicknesses under the condition of invariance of the total absorbed energy. The optimal structure configuration is established in terms of reducing the possible negative effect of laser impact on the pyroelectric coefficient stability.

Mathematical Modeling of the Cooling First Wall of the Tokamak Reactor

In the course of designing the post-ITER fusion facilities, a number of new technical and scientific problems occurred that were unknown before and necessarily have to be solved. Among these problems, one of the most important is the design of the reactor first wall, which could withstand the continuous heat fluxes coming from the fusion plasma. In this article, the mathematical model of the cooling first wall of the reactor is constructed and the numerical analysis is performed of the currently accepted first wall design of the future DEMO-FNS reactor (Fusion Neutron Source).

Plasma flow from vacuum surface flashover initiated by 70-kV 20-ns pulses

In this work we study plasma beams generated by vacuum surface flashover discharge. We measured ionic current, velocity of ions at values of discharge current 2.8 kA and 5.5 kA for polymethylmethacrylate (PMMA), polytetrafluorethylene (PTFE) and polyethylene (PE). The aim of the work is to compare parameters of particle flows at different values of discharge current and unchanged voltage. We used generators with coaxial and disk-shaped glycerol-filled pulse-forming lines. Maximum voltage is 70 kV; voltage pulse duration is 20 ns; width of a current pulse first half wave is 30 ns. Average velocity of the ions for all materials is within the range ∼ 100−150 km/s and doesn’t change significantly as the current increases from 2.8 kA to 5.5 kA. Ion current of the beam increase approximately linearly at higher discharge current. We estimate the density of electrons in plasma as a function of distance from the discharge. The applicability of this type of discharge for creating plasma for neutralizing high-power ion beams in inertial confinement fusion facilities is discussed.

Preliminary Reliability Allocation Method for Fusion Facility Based on Probabilistic Safety Assessment

The construction of ITER indicates that the development of fusion energy has entered the engineering stage. Reliability assurance, which is essential in the design and safety supervision of a new reactor, is an urgent problem to be solved in the process of fusion reactor engineering. But, up to now, the reliability work of a fusion device has not considered the security of the fusion reactor. This paper proposes a method to establish a reliability index for fusion safety–related components. The reliability index is a useful indicator to evaluate the reliability of the system, and it is also the regulatory basis for regulatory authorities. First, this paper gives the recommended values of probability safety goals for a fusion reactor. Second, the reliability requirements for the fusion safety system that meet the safety goals are calculated. In this part, a Probability Safety Assessment (PSA) is adopted to establish the risk models and calculate the undesired consequences of a fusion reactor. Based on the PSA analysis results, a risk-informed approach is used to categorize the plant structures, systems, and components of the fusion reactor as four categories according to their safety significance. Last, the reliability index of the safety-related components is given based on the results of the risk-informed safety categorization and PSA analysis results. The validation of the reliability index system is still being studied, and this work is expected to support the reliability evaluation and safety supervision of a fusion reactor in the future.

ITER Framework for RAMI Engineering

The systems engineering process starts with the discovery of the real issues that need to be resolved and the identification of failures that are the most probable or/and have the highest negative impact during the life cycle of a project. Systems engineering involves finding mitigations to these most critical problems. This logic is fully followed in reliability, availability, maintainability, inspectability (RAMI) engineering. Although this area is at its beginning in fusion technologies, a few years ago the ITER Organization developed an approach to assess the RAMI requirement of systems. As an example of what a RAMI analysis can bring to the maintainability and thus operational availability of a nuclear fusion facility like ITER, the availability of the cask and plug remote handling system in charge of handling of port plugs and their moving between the port cells to the hot cell facility is addressed in the case of diagnostic equatorial port plugs.

Bio-inspired surface aberration mitigation technique for high-frequency conversion efficiency of large-aperture nonlinear optics.

Low-frequency conversion efficiency severely limits the 3ω ultraviolet energy density at the target of inertial confinement fusion facilities. Here, we present a bio-inspired surface aberration mitigation (SAM) technique that could significantly reduce the crystal surface aberration and realize high frequency conversion efficiency over a clear aperture of 400 mm × 400 mm. Numerical models are utilized to optimize and verify the mechanical properties and physical performance of the SAM technique. In addition, the influences of various operation conditions on surface aberration, angle-detuning magnitude, and frequency conversion efficiency are illustrated. Finally, the process stability and online feasibility of this new approach are validated by the established offline characterization system and in situ imaging system.

Improvement of irradiation uniformity by dynamic interference structures of laser array in inertial-confinement-fusion facilities

In order to improve the irradiation uniformity in inertial confinement fusion facilities, a novel scheme, smoothing by interference structures of laser array, is firstly proposed for improving the irradiation characteristics. In this approach, a pair of identical phase plates with either tilted, cylindrical, elliptical or spiral phase modulates two beamlets, and another pair of phase plates that are conjugate with the former reshapes another two beamlets. Between the beamlets with conjugate wavefront distributions, their central wavelengths are shifted but their polarization states are the same. Arising from the combined effect of the conjugate wavefront distributions and the wavelength differences among the beamlets, the dynamic interference structures are generated in the focal region and thus improve the irradiation uniformity of these beams. Simulation shows that, the superposed focal spot of the beamlets in the focal region is smoothed in transverse, radial or azimuthal direction due to the phase types of the plates, and the required time to achieve the best uniformity is much shorter than the smoothing by spectral dispersion scheme. Moreover, the phase plates can be integrated with the continuous phase plates or be placed in the beamline as pure phase plates in the high-power laser facilities.

Optimization of MAXED input parameters with applications to the unfolding of neutron diagnostics data from the Joint European Torus.

Organic NE213 liquid scintillator neutron detectors are commonly used at accelerator facilities for neutron beam diagnostics. In recent years, they have also been installed at nuclear fusion facilities to measure the neutron energy spectra from Deuterium (D) and Deuterium-Tritium plasmas, e.g., at the ASDEX Upgrade (Garching, Germany) and at the Joint European Torus (JET, Culham, UK). The NE213 equivalent scintillating material (now BC501A) of the detector is sensitive to neutron and gamma radiation, so pulse discrimination techniques are applied in order to obtain the measured neutron pulse height spectrum (PHS). If the neutron detector is stable, controlled, and well-characterized (i.e., its response function to incoming neutrons of known energy is determined with high accuracy), it can be used as a neutron spectrometer. The measured PHS can then be analyzed using standard techniques such as unfolding to determine the incoming neutron energy spectrum. This article focuses on the unfolding of neutron data measured in D plasma experiments at JET by a compact broadband neutron spectrometer using the maximum entropy unfolding code MAXED. A general method for optimizing MAXED input parameters is described and applied to the measured PHS to diagnose the effects of the auxiliary heating of JET D plasma on the neutron energy spectra.

Time-stretched photonic Doppler velocimetry.

Inertial confinement fusion facilities generate implosions at speeds greater than 100 km/s, and measuring the material velocities is important and challenging. We have developed a new velocimetry technique that uses time-stretched spectral interferometry to increase the measurable velocity range normally limited by the detector bandwidth. In this approach, the signal is encoded on a chirped laser pulse that is stretched in time to reduce the beat frequency before detection. We demonstrate the technique on an imploding liner experiment at the Sandia National Laboratories' Z machine, where beat frequencies in excess of 50 GHz were measured with 20 GHz bandwidth detection.

Designing of cooling water system for a pyro-breaker utilized in superconductive fusion facility

Pyro-breaker used as a backup breaker in Quench Protection System (QPS) guarantees the safety of a superconductive fusion facility. High current flows through the pyro-breaker and great amount of heat flux will be brought into the breaker due to the structure and mechanism of the Pyro-breaker. Requirement of each subsystem is crucial for the overall designing of the entire Cooling Water System (CWS). In this paper an optimized design of the CWS for Pyro-breaker is presented. Theoretical calculation and Computational Fluid Dynamics (CFD) are used to analysis the heat transfer coefficient of the cooling system. Thermal-electrical simulation is performed on the model and the result is verified by experiment on a prototype. It can be concluded that the presented Pyro-breaker model has the current capacity of 40kA with the coolant velocity of 1 m/s. Optimization of the CWS is made based on the verified method and a set of required coolant velocities is proposed when the current capacity reaches up to 80kA. The pressure drop, which is an important factor for the entire CWS, is calculated in ANSYS CFX when the breaker is applied under 80kA.

Overview of the recent experimental research on the J-TEXT tokamak

Recent J-TEXT research has highlighted the significance of the role that non-axisymmetric magnetic perturbations, so called three-dimensional (3D) magnetic perturbation (MP) fields, play in a fundamentally 2D concept, i.e. tokamaks. This paper presents the J-TEXT results achieved over the last two years, especially on the impacts of 3D MP fields on magnetohydrodynamic instabilities, plasma disruptions and plasma turbulence transport.On J-TEXT, the resonant MP (RMP) system, capable of providing either a static or a high frequency (up to 8 kHz) rotating RMP field, has been upgraded by adding a new set of 12 in-vessel saddle coils. The shattered pellet injection system was built in J-TEXT in the spring of 2018. The new capabilities advance J-TEXT to be at the forefront of international magnetic fusion facilities, allowing flexible study of 3D effects and disruption mitigation in a tokamak.The fast rotating RMP field has been successfully applied for avoidance of mode locking and the prevention of plasma disruption. A new control strategy, which applies pulsed RMP to the tearing mode only during the accelerating phase region, was proved by nonlinear numerical modelling to be efficient in accelerating mode rotation and even completely suppresses the mode. Remarkably, the rotating tearing mode was completely suppressed by the electrode biasing. The impacts of 3D magnetic topology on the turbulence has been investigated on J-TEXT. It is found that the fluctuations of electron density, electron temperature and plasma potential can be significantly modulated by the island structure, and a larger fluctuation level appears at the X-point of islands. The suppression of runaway electrons during disruptions is essential to the operation of ITER, and it has been reached by utilizing the 3D magnetic perturbations on J-TEXT. This may provide an alternative mechanism of runaway suppression for large-scale tokamaks and ITER.

Application of ADVANTG to the JET3 – NEXP streaming benchmark experiment

An application of the ADVANTG (AutomateD VAriaNce reducTion Generator) code, which combines the Denovo deterministic transport solver with the MCNP (Monte Carlo N-Particle) Monte Carlo code, to the JET3-NEXP (Joint European Torus work package 3) streaming benchmark experiment is presented in this paper. An ADVANTG input parameter variation analysis was performed in order to find optimal input parameters for the hybrid two-step workflow. ADVANTG-accelerated calculations from three different institutions (JSI (Jožef Stefan Institute), ORNL (Oak Ridge National Laboratory), and CCFE (Culham Centre for Fusion Energy)) were compared to analog MCNP simulations confirming no bias is introduced due to the use of ADVANTG. Additionally, ADVANTG-accelerated MCNP numerical simulations of the neutron fluence were compared to experimental results performed in 2016 at JET using thermo-luminescence detectors. Calculation/Experiment values from 0.7 to 13 were calculated for the experimental positions in the SW (southwest) labyrinth and SE (southeast) chimney. Using ADVANTG-generated variance reduction parameters, speed-up factors of up to 1100 relative to analog calculations were achieved. The MCNP statistical tests on track length estimator volume averaged tallies were consistently passed for all experimental locations with ADVANTG-generated variance reduction parameters. These results demonstrate that ADVANTG is capable of accelerating tally convergence at locations far from the source in streaming-dominated transport simulations of complex fusion facility models.