Total Fusion Power Research Articles

Overview of interpretive modelling of fusion performance in JET DTE2 discharges with TRANSP

In the paper we present an overview of interpretive modelling of a database of JET-ILW 2021 D-T discharges using the TRANSP code. The main aim is to assess our capability of computationally reproducing the fusion performance of various D-T plasma scenarios using different external heating and D-T mixtures, and to understand the performance driving mechanisms. We find that interpretive simulations confirm a general power-law relationship between increasing external heating power and fusion output, which is supported by absolutely calibrated neutron yield measurements. A comparison of measured and computed D-T neutron rates shows that the calculations’ discrepancy depends on the absolute neutron yield. The calculations are found to agree well with measurements for higher performing discharges with external heating power above ∼20 , while low-neutron shots display an average discrepancy of around +40% compared to measured neutron yields. A similar trend is found for the ratio between thermal and beam-target fusion, where larger discrepancies are seen in shots with dominant beam-driven performance. We compare the observations to studies of JET-ILW D discharges, to find that on average the fusion performance is well modelled over a range of heating power, although an increased unsystematic deviation for lower-performing shots is observed. The ratio between thermal and beam-induced D-T fusion is found to be increasing weakly with growing external heating power, with a maximum value of 1 achieved in a baseline scenario experiment. An evaluation of the fusion power computational uncertainty shows a strong dependence on the plasma scenario type and fusion drive characteristics, varying between ±25% and 35%. D-T fusion alpha simulations show that the ratio between volume-integrated electron and ion heating from alphas is 10 for the majority of analysed discharges. Alphas are computed to contribute between ∼15% and 40% to the total electron heating in the core of highest performing D-T discharges. An alternative workflow to TRANSP was employed to model JET D-T plasmas with the highest fusion yield and dominant non-thermal fusion component because of the use of fundamental radio-frequency heating of a large minority in the scenario, which is calculated to have provided ∼10% to the total fusion power.

Neutronic analysis and measurement performance assessment of ITER neutron flux monitor system

The ITER Neutron Flux Monitor (NFM) system consists of four subsystems located at ITER Equatorial Ports 1#, 7#, 8#, and 17#. NFM system carries out the measurement of total neutron yield and fusion power by employing 235U fission chambers. The total neutron yield to be measured spans a range of 7 orders of magnitude — from 1014 to 3.2 × 1020 n/s. Suitable moderator and detectors were selected to cover such a wide range of neutron yield and meet the requirements of time resolution and accuracy. Moderators and detectors’ sensitivities of NFM system are reasonably designed. The detectors’ sensitivities of the NFM#01 system under different moderators have been calculated with MCNP code, and graphite serves as the optimal moderator. The detectors’ sensitivities of the other three NFM subsystems have also been simulated simultaneously. Based on the required sensitivities of detector, the appropriate mass of 235U was selected for each fission chamber. According to the analysis of the measurement range of each NFM subsystem, it is finally confirmed that the measurement performance of the NFM system can meet the requirements well.

Design of ITER neutron flux monitor system back-end electronics

The primary aim of the International Thermonuclear Experimental Reactor (ITER) Neutron Flux Monitor (NFM) system is to diagnose the total neutron emission rate and fusion power of the plasma fusion reaction. The neutron flux is measured directly by the system and the fusion power can be calculated according to the neutron flux result. The NFM system is composed of detectors and back-end electronics. This paper focuses on the work of NFM#7 back-end electronics design and development. The back-end electronics consist mainly of four parts: Firstly, the pre-amplifiers are used to charge integrate and amplify the signal of fission chambers. Secondly, the output of pre-amplifiers will be transmitted to Signal Processing Unit (SPU) for analog signal conditioning, data processing, and measurement result uploading. Thirdly, pre-amplifiers and fission chambers have different electricity sources. Pre-amplifiers receive electricity generated from low voltage power whilst fission chambers receive electricity generated from high voltage power. Lastly, the Instrument and Control (I&C) is designed as the interface between NFM hardware and ITER central system and its functions are configuration & control, monitoring, calibration, data communication & archiving and clock synchronization. All the four parts are well integrated and tested in neutron field. The system design also follows the guidelines and standards of ITER.

Data Acquisition System for ITER Neutron Diagnostic Divertor Neutron Flux Monitor

Neutron diagnostic “Divertor Neutron Flux Monitor” is being under the development for the ITER project. This diagnostic is a primary tool for total neutron yield and fusion power measurements. Diagnostic consists of the three detector modules and the enhanced data acquisition (DAQ) systems. Each module contains three fission chambers (FCs) with 235U and three with 238U. The choice of the FC types and sensitivities is determined by the project requirements for the neutron flux measurement range, error, and time resolution. The DAQ system performs preliminary amplification and processing of the fission chamber signals, and the calculation of the neutron flux at the module location to calculate the total neutron yield and fusion power. The DAQ system transfers measured data to the upper level of the control system with a time resolution of 1 ms and provides raw data archiving. The hardware and software of the DAQ system were developed taking into account the requirements of the ITER project. The data processing system mockup of a single detection module was manufactured. This mockup passed laboratory and reactor tests. The system has passed the preliminary design approval. Preparation for the manufacturing of the system prototype is carried out at this moment. The architecture of the DAQ system and its components are shown in the report, and the test results of the system mockup are discussed.

Effects of kinetic profiles on neutron wall loading distribution in CFETR

The China Fusion Engineering Test Reactor (CFETR) is under design, which aims to bridge the gaps between ITER and the future fusion power plant. The neutron wall loading (NWL) depends on the neutron source distribution, which depends on the density and temperature profiles. In this paper, we calculate the NWL of CFETR and study the effects of density and temperature profiles on the NWL distribution along the first wall. Our calculations show that for a 200 MW fusion power, the maximum NWL is at the outer midplane and the vaule is about 0.4 MW m−2. The density and temperature profiles have little effect on the NWL distribution. The value of NWL is determined by the total fusion power.

Neural network implementation for ITER neutron emissivity profile recognition

The ITER Radial Neutron Camera (RNC) is a neutron diagnostic intended for the measurement of the neutron emissivity radial profile and the estimate of the total fusion power. This paper presents a proof-of-principle method based on neural networks to estimate the neutron emissivity profile in different ITER scenarios and for different RNC architectures. The design, optimization and training of the implemented neural network is presented together with a decision algorithm to select, among the multiple trained neural networks, which one provides the inverted neutron emissivity profile closest to the input one. Examples are given for a selection of ITER scenarios and RNC architectures. The results from this study indicate that neural networks for the neutron emissivity recognition in ITER can achieve an accuracy and precision within the spatial and temporal requirements set by ITER for such a diagnostic.

Neutron Detector Needs for ITER

For a long time, nuclear fusion has been anticipated to become a future power source. The International Thermonuclear Experimental Reactor (ITER) tokamak is designed to demonstrate the feasibility of applying the deuterium-tritium fusion reaction to human power needs. The measurements of ITER's fusion neutron flux parameters can provide information on total fusion power and fusion power density as well as other plasma parameters. This paper gives an overview of the technical constraints in terms of the radiological, thermal, and electromagnetic loads for ITER neutron detectors. These constraints have been studied and summarized with measurement requirements. The areas of high risk have been highlighted to encourage research and development of neutron detectors for the urgent needs of ITER.

Fusion–Fission System Based on a Tokamak with Strongly Radiative Edge

The current manuscript presents a conceptual analysis of the hybrid tokamak operating with D–T plasmas in regimes with high values of the fusion gain Q (Q = fusion power output/auxiliary heating power input) or even in the ignition regime. The total fusion power release can reach \(P_{\text {fus}}\) = 150–200 MW. The key features of the hybrid system fusion–fission are a relatively small size and a construction simplicity of this plasma device (minor and major radii of the tokamak are \(a = 1.2\) m and \(R = 4.4\) m, respectively). It is assumed that the hybrid tokamak will operate in the so-called radiative improved (RI-)mode which has been extensively investigated in the tokamak TEXTOR-94 and other machines. This regime allows to overcome a range of practical difficulties existing nowadays to tokamak-reactors (for instance, the power exhaust). Additionally, the RI-mode gives a possibility to extend the operational limits, in particular, to increase considerably the plasma density. The main parameters of the designed tokamak have been obtained using a simple system code. The critical issues of using such a tokamak as a basis for a fusion neutron source or for a hybrid reactor device are discussed.

Diagnostic neutron activation system for KSTAR

The Korea Superconducting Tokamak Advanced Research (KSTAR) device has begun deuterium plasma operation and increased neutron generation from D-D fusion reaction in the plasma is expected as the heating power of the plasma increases. A neutron activation system utilizing the pneumatic transfer of encapsulated metal samples has been implemented to monitor the fusion neutron source strength and the total fusion power from the KSTAR plasma. The prototype pneumatic transfer system for the ITER neutron activation system was slightly modified to be used for KSTAR, and a High Purity Germanium (HPGe) detection system was used to count gamma photons from the activated samples. The Monte Carlo code MCNP and the inventory code FISPACT were also used for the calculations to evaluate the total number of neutrons emitted from the D-D fusion reactions in the KSTAR plasma. The analysis of data from an activation measurement obtained from the 2011 KSTAR campaign shows the neutron flux at the irradiation station was 2.2 × 108 cm−2s−1, and the total neutron yield was 4.7 × 1013 n/s for a typical NBI-heated, H-mode KSTAR plasma shot.

Effect of pedestal height and internal transport barriers on International Thermonuclear Experimental Reactor target steady state simulations

The Tokamak simulation code (TSC) is used to provide initial conditions for predictive TRANSPort and integrated modeling code (PTRANSP) simulations of ITER target steady state scenarios. The PTRANSP simulations are carried out using the new multi-mode (MMM7.1) and the gyro-Landau-fluid (GLF23) transport models. It is found that there are circumstances under which the total fusion power decreases with increasing pedestal temperature height. When the total current (from magnetic axis to plasma edge) is fixed, an increased fraction of the current is concentrated in the pedestal region as the pedestal height is increased. As a consequence of the fixed total current, this results a smaller fraction of the current in the core plasma and, consequently, lower energy confinement. In previous simulations of ITER, in which the fusion power increased with increasing pedestal temperature height, the plasma current from the top of the pedestal to the magnetic axis was held fixed independent of the pedestal temperature. Simulations presented in this paper also indicate that improvement in fusion power production occurs when the lower hybrid current drive is replaced with electron cyclotron current drive. Again, the improvement results from the redistribution of plasma current since the lower hybrid power generally drives current closer to the plasma edge than does the electron cyclotron power. ITER simulation results obtained using the MMM7.1 transport model are compared with those using the GLF23 model. It is found that, in simulations of target steady state scenarios, momentum transport and flow-shear suppression features of the new MMM7.1 model can lead to predictions of internal transport barriers in temperature and rotation frequency.

Relationship between neutron yield rate of tokamak plasmas and spectrometer measured flux for different sight lines

A parametric relationship between total neutron yield rate and collimated fluxes related to the brightness (B) of plasma chords (λ) is developed for different emissivity distributions of tokamak plasmas. Specifically, the brightness was expressed as a function of chord coordinates of radial position using a simple model for the emissivity profiles of width parameter w. The functional brightness dependence B(λ,w) was calculated to examine the relationship between measured flux and deduced yield rate, and its plasma profile dependence. The results were used to determine the chord range of minimum profile sensitivity in order to identify the preferred collimator sight for the determination of yield rate from neutron emission spectroscopy (YNES) measurements. The YNES method is discussed in comparison to conventional methods to determine the total neutron yield rates and related plasma fusion power relying on uncollimated flux measurements and a different calibration base for the flux-yield relationship. The results have a special bearing for tokamaks operating with both deuterium and deuterium-tritium plasmas and future high power machines such as for ITER, DEMO, and IGNITOR.

Operational Radioactivity Evaluation of ITER Diagnostic Neutron Activation System

At the International Thermonuclear Experimental Reactor (ITER), a neutron activation system is foreseen to measure neutron fluence on the first wall and to evaluate total fusion power from ITER plasma. This system utilizes the method of counting gamma radiation from the metal sample irradiated by the neutron flux at the irradiation end located near the plasma. Evaluation of the operational radioactivity of the sample, which has been induced during ITER operation, is necessary in order to determine sample material, mass, and the appropriate irradiation time. Neutronic calculations are performed to evaluate operational activity of the sample. Activation of the sample is calculated according to the irradiation time for various sample materials such as Si, Al, Ti, Fe, Nb, and Cu and by assuming that the samples are irradiated at the midplane inboard region where the neutron flux is one of the strongest.

In-vessel design of ITER diagnostic neutron activation system

A neutron activation system utilizing an encapsulated metal sample technique will be implemented to evaluate total fusion power from ITER plasma. The positions of seven irradiation ends for each poloidal section are determined to optimize the measurements from asymmetric, elongated plasma with minimum uncertainty. To determine the exact location of the irradiation end, some physical and technical aspects for each location are investigated. Monte Carlo neutronic calculations are performed to assess the physical characteristics and optimize the design of each irradiation end.

On the extrapolation to ITER of discharges in present tokamaks

An expression for the extrapolated fusion gain G = Pfusion/5Pheat (Pfusion being the total fusion power and Pheat the total heating power) of ITER in terms of the confinement improvement factor (H) and the normalized beta (βN) is derived in this paper. It is shown that an increase in normalized beta can be expected to have a negative or neutral influence on G depending on the chosen confinement scaling law. Figures of merit like should be used with care, since large values of this quantity do not guarantee high values of G, and might not be attainable with the heating power installed on ITER.

Conceptual design of an evaporation-cooled liquid metal divertor for fusion power plants

In fusion power plant divertors a significant portion of the total fusion power is absorbed. In order to achieve a high thermal plant efficiency this power has to be removed at high temperature levels. This calls for innovative designs which differ from those proposed for experimental fusion machines. In this study, evaporation cooling of liquid metals is considered evaluating: (i) a concept based on pool boiling; and (ii) a concept based on the transport of liquid through a capillary layer and evaporation at the surface of this layer. The latter concept employs characteristic features of high heat flux liquid metal heat pipes. Both concepts are characterised by the operation at high temperatures (between 750 and 1200 °C, depending on the choice of the liquid metal) but small stresses due to small temperature differences and operating pressures. The second concept is assessed to have the potential of significantly higher heat loads (in space craft applications heat loads of much larger than 10 MW/m 2 were reached). A conceptional design description is presented; divertor relevant performance limits are estimated and open issues and the required R&D work are briefly addressed.

Neoclassical impurity transport in the core of an ignited tokamak plasma

Impurity behaviour in the core of an ignited fusion reactor(ITER FDR) is investigated under the assumption that theimpurity fluxes are dominated by neoclassical transport.Equilibrium impurity profiles are evaluated for twotemperature profiles and compared with the resultsfor purely anomalous diffusion with a core diffusioncoefficient of Dan = 0.5 m2/s.The neoclassical fluxes are calculated taking collisions of all relevantplasma species into account.The diffusion coefficient in the core is below 0.1 m2/sand decreases with increasing Z of the impurity. The mean driftvelocities are always outwardly directed and cause an effective screeningfor high-Z elements.Due to the low diffusion coefficient, the He concentration profile in the core isstrongly rising. Helium concentrations on-axis are cHe ≈ 15-17%. In contrast to the calculation with purely anomalous transport andcHe ≈ 9% on-axis,the He density is not only determined by the recycling edge source butby the interplay of low core transport, increasing dilution and decreasing fusion source. The increased dilution causes a reduction of total fusion power by ≈ 8-11%.With an additional anomalous core diffusion for He of DanHe = 0.2 m2/s the central He concentration decreases to cHe ≈ 10%.

Use of polarized helium-3 for the energy production

After the discovery of plentiful of minable helium-3 on the lunar surface, fusion scientists have tried to find out the methods for utilizing this attractive fuel for fusion reactors. Among various approaches, a field-reversed configuration (FRC) for confinement of fusion plasma appears the most promising candidate for a D- 3He fueled fusion reactor. A conceptual design of D- 3He fueled FRC reactor “ARTEMIS” (Momota et al., in: Proc. 7th Int. Conf. on Emerging Nuclear Energy Systems, Chiba, Japan, 1993) has been carried out showing that bases of engineering needed for achieving a commercial reactor are conventional and the cost of electricity from “ARTEMIS” is estimated as cheap as approximately 30 mills/(kWh). A low neutron yield allows us large freedom of reactor materials and reduces the problem of disposal of radioactive waste. In this paper, we will examine the use of polarized D- 3He fuels. A favorable characteristic of polarized fuels lies on the reduction of neutron yield and the compactness of the fusion core. Because of the enhancement of reactivity of D- 3He reaction by applying polarization, the neutron fraction in the total fusion power decreases to 1.3% with the same volume of “ARTEMIS” and consequently neutron wall loading is 72 kW/m 2. The wall loading should be compared with 180 kW/m 2 from “ARTEMIS” or 10 000 kW/m 2 from a D-T fueled reactor. Consequently, this low neutron mode gives us large freedom in choosing reactor materials. The fusion reactor using polarized fuels with the same neutron yields as “ARTEMIS” decreases its size from 196 to 33 m 3. This operation mode gives us the possibility of developing an economic fusion reactor.

Review of deuterium–tritium results from the Tokamak Fusion Test Reactor

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

Design of the ITER vacuum vessel

The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement. The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2—2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4−0.7 m. The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200–300 MN due to VDE and to allow for differential thermal expansion. The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology.

Evaluation of Current Drive Requirements and Operating Characteristics of a High Bootstrap Fraction Advanced Tokamak Reactor

The reactor potential of some advanced physics operating modes proposed for the TPX physics program [1] are examined. A moderate aspect ratio (A = 4.5 as in TPX), 2 GW reactor (see Table I for parameters) is analyzed because of its potential for steady-state, non-inductive operation with high bootstrap current fraction. Particle, energy and toroidal current equations are evolved to steady-state conditions using the 1-1/2-D time-dependent WHIST transport code [2]. The solutions are therefore consistent with particle, energy and current sources and assumed transport models. Fast wave current drive (FWCD) provides the axial seed current. The bootstrap current typically provides 80–90% of the current, while feedback on the lower hybrid current drive (LHCD) power maintains the total current. The sensitivity of the plasma power amplification factor, Q ≡ Pfus/Paux, to variations in the plasma properties is examined. The auxiliary current drive power, Paux = PLH + PFW; bootstrap current fraction; current drive efficiency; and other parameters are evaluated. The plasma is thermodynamically stable for the energy confinement model assumed (a multiple of ITER89P). The FWCD and LHCD sources provide attractive control possibilities, not only for the current profile, but also for the total fusion power since the gain on the incremental auxiliary power is typically 10–30 in these calculations when overall Q ≍ 30.