Power Multiplication Factor Research Articles

MANTA: a negative-triangularity NASEM-compliant fusion pilot plant

Abstract The MANTA (Modular Adjustable Negative Triangularity ARC-class) design study investigated how negative-triangularity (NT) may be leveraged in a compact, fusion pilot plant (FPP) to take a ‘power-handling first’ approach. The result is a pulsed, radiative, ELM-free tokamak that satisfies and exceeds the FPP requirements described in the 2021 National Academies of Sciences, Engineering, and Medicine (NASEM) report ‘Bringing Fusion to the U.S. Grid’ (2021 Bringing Fusion to the U.S. Grid). A self-consistent integrated modeling workflow predicts a fusion power of 450 MW and a plasma gain of 11.5 with only 23.5 MW of power to the scrape-off layer (SOL). This low P SOL together with impurity seeding and high density at the separatrix results in a peak heat flux of just 2.8 MW m−2. MANTA’s high aspect ratio provides space for a large central solenoid (CS), resulting in ∼15 minute inductive pulses. In spite of the high B fields on the CS and the other REBCO-based magnets, the electromagnetic stresses remain below structural and critical current density limits. Iterative optimization of neutron shielding and tritium breeding blanket yield tritium self-sufficiency with a breeding ratio of 1.15, a blanket power multiplication factor of 1.11, toroidal field coil lifetimes of 3100 ± 400 MW · yr, and poloidal field coil lifetimes of at least 890 ± 40 MW · yr. Following balance of plant modeling, MANTA is projected to generate 90 MW of net electricity at an electricity gain factor of ∼ 2.4 . Systems-level economic analysis estimates an overnight cost of US$3.4 billion, meeting the NASEM FPP requirement that this first-of-a-kind be less than US$5 billion. The toroidal field coil cost and replacement time are the most critical upfront and lifetime cost drivers, respectively.

Scaling laws from theory-based modeling for different regimes in the DEMO fusion reactor

In the European Union strategy, DEMO is the single step between ITER and a fusion power plant. The current state of physical knowledge and technological limits does not yet well allow us to identify the best combination of solutions to decide the architecture of this new tokamak. To this purpose, an important element of tokamak research finalized to the prediction, the interpretation and the planning of experiments is represented by the ‘integrated modeling’. In this work we use the integrated modeling ASTRA tool coupled with the TGLF transport model in order to investigate scenarios of interest for DEMO. We obtain scaling laws which describe and serve as a guide-line for the pre-conceptual design of the machine. In particular, by means of scaling laws we investigate the performances of the reactor in terms of fusion power, confinement time H-factor and fusion power multiplication factor Q, in different regimes such as H-mode, I-mode and L-mode. In this way we provide important key points for the establishment of the best physical and technological solutions for DEMO scenarios.

Investigation of key parameters for the development of reliable ITER baseline operation scenarios using CORSICA

ITER will demonstrate the feasibility of burning plasma operation by operating DT plasmas in the ELMy H-mode regime with a high ratio of fusion power gain Q ~ 10. 15 MA ITER baseline operation scenario has been studied using CORSICA, focusing on the entry to burn, flat-top burning plasma operation and exit from burn. The burning plasma operation for about 400 s of the current flat-top was achieved in H-mode within the various engineering constraints imposed by the poloidal field coil and power supply systems. The target fusion gain (Q ~ 10) was achievable in the 15 MA ITER baseline operation with a moderate amount of the total auxiliary heating power (~50 MW). It has been observed that the tungsten (W) concentration needs to be maintained low level (nw/ne up to the order of 1.0 × 10−5) to avoid the radiative collapse and uncontrolled early termination of the discharge. The dynamic evolution of the density can modify the H-mode access unless the applied auxiliary heating power is significantly higher than the H-mode threshold power. Several qualitative sensitivity studies have been performed to provide guidance for further optimizing the plasma operation and performance. Increasing the density profile peaking factor was quite effective in increasing the alpha particle self-heating power and fusion power multiplication factor. Varying the combination of auxiliary heating power has shown that the fusion power multiplication factor can be reduced along with the increase in the total auxiliary heating power. As the 15 MA ITER baseline operation scenario requires full capacity of the coil and power supply systems, the operation window for H-mode access and shape modification was narrow. The updated ITER baseline operation scenarios developed in this work will become a basis for further optimization studies necessary along with the improvement in understanding the burning plasma physics.

CORSICA modelling of ITER hybrid operation scenariosbbExtended from Proc. 24th Int. Conf. on Fusion Energy (San Diego, 2012) IT/P1-13.

The hybrid operating mode observed in several tokamaks is characterized by further enhancement over the high plasma confinement (H-mode) associated with reduced magneto-hydro-dynamic (MHD) instabilities linked to a stationary flat safety factor () profile in the core region. The proposed ITER hybrid operation is currently aiming at operating for a long burn duration (>1000 s) with a moderate fusion power multiplication factor, , of at least 5. This paper presents candidate ITER hybrid operation scenarios developed using a free-boundary transport modelling code, CORSICA, taking all relevant physics and engineering constraints into account. The ITER hybrid operation scenarios have been developed by tailoring the 15 MA baseline ITER inductive H-mode scenario. Accessible operation conditions for ITER hybrid operation and achievable range of plasma parameters have been investigated considering uncertainties on the plasma confinement and transport. ITER operation capability for avoiding the poloidal field coil current, field and force limits has been examined by applying different current ramp rates, flat-top plasma currents and densities, and pre-magnetization of the poloidal field coils. Various combinations of heating and current drive (H&CD) schemes have been applied to study several physics issues, such as the plasma current density profile tailoring, enhancement of the plasma energy confinement and fusion power generation. A parameterized edge pedestal model based on EPED1 added to the CORSICA code has been applied to hybrid operation scenarios. Finally, fully self-consistent free-boundary transport simulations have been performed to provide information on the poloidal field coil voltage demands and to study the controllability with the ITER controllers.

PARCS code multi-group neutron diffusion constants generation using Monte Carlo method

PARCS code is a three-dimensional (3D) reactor core simulator which solves the steady-state and time-dependent multi-group neutron diffusion equations if the multi-group diffusion constants (MGDCs) are provided. The MGDCs are mostly prepared for reactor physics problems using deterministic lattice codes. Beside approximation in the geometry, a lattice code inherently applies estimates to the neutron transport model. On the other hand, the geometric flexibility and use of continuous energy cross sections data library associated with the Monte Carlo (MC) method makes it a good candidate for the generation of highly accurate multi-group cross sections. In this study, a new MC based methodology is applied to generate the MGDCs which can be utilized in the PARCS code input file directly or as PMAXS files for a reactor core simulation. To achieve this, a new tool in MATLAB software is developed to compute the MGDCs from the MCNPX 2.7 MC code outputs. Verification of the proposed method for two-group constants generation is carried out using Tehran research reactor (TRR) core simulation in different steady state conditions. The calculated values of axial and radial power distributions and multiplication factor using the PARCS code are verified against the MCNPX 2.7 code results. The results illustrate that the proposed method has high accuracy in MGDCs generation.

LDPC codes for soft decode-and-forward in half-duplex relay channels

We investigate the use of rate-compatible lowdensity parity-check (RC-LDPC) codes as part of a soft decodeand- forward (SDF) protocol over the half-duplex relay channel. We propose a new methodology to design the degree distribution of the RC-LDPC codes with a lower triangular parity-check matrix, enabling the additional parity bits to be linearly and systematically encoded at the relay. Our proposed methodology introduces the concept of a K-layer doping matrix to represent the structure of a lower triangular parity-check matrix. As a result of our methodology, the asymptotic performance of RC-LDPC codes can be analyzed and predicted using the multi-edge-type density evolution. Then, we derive the soft-re-encoding of the additional parity symbols at the relay using our designed RCLDPC codes. Moreover, we propose a novel method, which we refer to as soft fading, to compute the log-likelihood ratio (LLR) of the received signal at the destination for the SDF protocol. We demonstrate that our proposed soft fading method outperforms the best known method in the literature by up to 0.7 dB in terms of BER performance. Finally, we derive a new bound for the power multiplication factor at the relay, which limits the amount of soft-errors forwarded by the relay to the destination. The BER performance of our new RC-LDPC codes improves significantly once the power multiplication factor at the relay satisfies this bound.

LDPC codes for soft decode-and-forward in half-duplex relay channels

We investigate the use of rate-compatible lowdensity parity-check (RC-LDPC) codes as part of a soft decodeand- forward (SDF) protocol over the half-duplex relay channel. We propose a new methodology to design the degree distribution of the RC-LDPC codes with a lower triangular parity-check matrix, enabling the additional parity bits to be linearly and systematically encoded at the relay. Our proposed methodology introduces the concept of a K-layer doping matrix to represent the structure of a lower triangular parity-check matrix. As a result of our methodology, the asymptotic performance of RC-LDPC codes can be analyzed and predicted using the multi-edge-type density evolution. Then, we derive the soft-re-encoding of the additional parity symbols at the relay using our designed RCLDPC codes. Moreover, we propose a novel method, which we refer to as soft fading, to compute the log-likelihood ratio (LLR) of the received signal at the destination for the SDF protocol. We demonstrate that our proposed soft fading method outperforms the best known method in the literature by up to 0.7 dB in terms of BER performance. Finally, we derive a new bound for the power multiplication factor at the relay, which limits the amount of soft-errors forwarded by the relay to the destination. The BER performance of our new RC-LDPC codes improves significantly once the power multiplication factor at the relay satisfies this bound.

Inductively Operated Day-Long Tokamak Reactor with High Power Multiplication Factor.

Tokamak D-T power reactor with day-long operation is investigated, where plasma current is driven inductively.Thereby the high power multiplication factor (typically Q>1, 000) is attained easily.For day-long operation only with the use of inductive current drive, the large plasma major radius is necessary to provide sufficient magnetic flux in the Ohmic transformer.Parameters of tokamak reactor for day-long operation are typically major radius 12m, minor radius 1.5m with elongation ratio1.6 and triangularity0.5, toroidal mag netic field12.5T, and plasma current 8.7MA.Plasma volume is comparable to the current-driven reactors, such as SSTR and ARIES in spite of large major radius, because the aspect ratio is relatively high.

D-3He fusion yield with higher harmonic ICRF heating of 3He beams

The enhancement of the D-3He fusion yield by combined 3He beam injection and higher harmonic ICRF heating is investigated on the basis of the local Fokker-Planck calculation. With the fourth harmonic heating of 100 keV 3He beams the fusion yield is enhanced most efficiently when the ratio of the beam power to the radiofrequency power is 1/5. The maximum of the fusion power multiplication factor is roughly given by ΔQmax ∼ 0.043(Te[10 keV])(ne [1020 m−3])1.5. A comparison is made with the case of fundamental minority heating of 3He beams and the case of injection of 500 keV D beams.

Improvement of fusion reactivity and fusion power multiplication factor in the presence of fast ions

Improvement of fusion reactivity (〈σv〉) and fusion power multiplication factor ( Q) in the presence of fast ions has been investigated with the analytical solution of the velocity distribution function. Fast ions created by only NBI heating, only ICRF heating and a combination of NBI and ICRF heating are considered in a D-T plasma. The dependences of 〈σv〉 on the plasma temperature, the deposited power density, the electron density and the injected beam energy are clarified. Calculations of evaluating Q employ a simple plasma model in which the radial profiles of the plasma temperature and density are the same as those observed in recent experiments. 〈σv〉 and Q are enhanced by only NBI heating, the combined NBI and ICRF heating and only ICRF heating in this order under the same heating power. However, the increase in 〈σv〉 and Q for the former two methods is almost the same. In the presence of fast ions, the enhancement of 〈σv〉 and Q is larger than the increase in the plasma pressure.

Impact of neutronics considerations on the selection of solid breeder and multiplier materials and configurations

A large number of solid breeder materials have been proposed for fusion blankets. A study was performed to quantitatively compare the effect of solid breeder (SB) material choice on blanket attractiveness and to recommend priorities for focusing the solid breeder experiments. This paper summarizes the results of the extensive neutronics analysis which evaluated the achievable tritium breeding ratio (TBR) and the power multiplication factor (M) as key parameters. The study considered ten solid breeder materials (e.g., Li2O, Li2SiO3, LiAlO2), six neutron multipliers (Be, BeO, Pb, PbO, PbBi and and Zr5Pb3) and two coolants (helium and water). These breeder and multiplier materials were considered in six different design configurations which included homogenized breeder/multiplier mixtures and a separate multiplier zones. The general conclusions which can be made from the TBR results are that: (1) Be exhibits the most superior performance among neutron multiplier (M) candidates; (2) Li2O is the most attractive solid breeder candidate, showing the highest TBR for the unmultiplied, separate multiplier and homogeneous SB/M cases; (3) Li2O is the only solid breeder with a plausible chance of performing satisfactorily without a multiplier; (4) the homogeneous SB/Be cases show exceptional TBR performance and should be considered in the test program, (5) multi-region solid breeder multiplier (SB/M) cases show still excellent TBR performance; and (6) the TBR with water coolant in all cases is 5–10% higher than with helium coolant. The power multiplication results tend to vary in accordance with the TBR results and thus reinforce the above conclusions. There are two notable exceptions; the first is the case of Li7Pb2, which, although showing a high effective TBR of 1.22, exhibits a rather low M of 1.18. The other exception is that changes in 6Li enrichment affect M only slightly in all cases. The overall M is actually decreased due to the decrease of the radiative capture reaction rate in the structure material used in the blanket utilizing highly enriched lithium.

Enhancement of fusion power multiplication factor

Enhancement of the fusion power multiplication factor, Q , has been investigated using a simple plasma model, in which the radial profiles of the plasma temperature and density or the fast-ion velocity distribution are the same as those observed in recent experiments. The break-even condition is relaxed by a peaked density profile, beam-plasma reactions, and the fast ions produced by ICRF heating at the second harmonic of deuterium in a D-T thermal plasma. The minimum n τ T required for break-even has also been lowered. Furthermore, it is possible to enhance Q through local thermal instability near the plasma center using a simple transport simulation, in which the profile of the plasma density does not change during the simulation and only the thermal conduction loss is considered. Finally, the operational plasma parameters for JT-60 in the n τ vs. T diagram have been investigated.

Scaling of the thermonuclear fusion neutron yield in the Doublet III tokamak

The fusion neutron yield during hydrogen neutral beam injection into deuterium plasmas is examined and compared with the neoclassical model, using the Doublet III data obtained in 1983. This provides scalings of thermonuclear fusion neutron yields for the discussion of ion energy transport in tokamak devices. – Experimental data show that fusion neutron yield appears to have a scaling of in low recycling divertor discharges (so called H mode). The neoclassical theory for ion heat conduction satisfactorily described the scaling of fusion neutron yield. With a neutral injection power of Pinj = 4.6 MW, the Doublet III deuterium plasma yields thermonuclear neutrons at a rate of 1.2 × 1013 n·s−1 in low recycling divertor discharge, which is equivalent to a D+ T+ plasma with a thermonuclear fusion power multiplication factor of Q = 7.8 × 10−5 or a beam driven T+ plasma with Q = 8.6 × 10−2. A limiter discharge yields fewer neutrons.

Tandem Mirror Reactor Power Balance Studies

A tandem mirror reactor (TMR) power plant balance model has been developed and is now being used as a computer aid for performing parametric studies. End-cell power injection into the plasma and the physics thermal Q are used to determine the fusion power. About 80% of the fusion power is transferred by high-energy neutrons to the blanket modules and structures. The other 20% of the fusion power in the high-energy alpha particles is used to heat the deuterium-tritium (D-T) plasma. Most of the plasma-ionized particles transfer their energy to the halo dumps and direct converters. The plant efficiency is calculated for three different system cycles: (1) the pressurized water/saturated steam cycle; (2) the superheated steam cycle; and (3) the more complex superheat/reheat cycle. There is a signficiant improvement in plant efficiency as the electrical power multiplication factor and steam cycle efficiency increases.

Thermonuclear Burning in a Tokamak Plasma with Polarized Nuclei

The regimes of thermonuclear burning in selfsustained and driven tokamak reactors using deuterium-tritium plasma with nuclei polarized along the magnetic field are investigated. A comparison is made between the burning regimes in reactors with polarized and unpolarized plasma. In particular, it is shown that the temperature regions that allow stable steady-state thermonuclear reactions are similar for both types of reactors. However, as compared to the conventional case, the driven reactor with polarized nuclei requires higher power levels of neutral injection or radiofrequency heating to achieve the same stable temperature regime. The power multiplication factor, when using polarized nuclei, is unchanged or may be higher due to deterioration of alpha-particle confinement.

Princeton Large Tokamak Scaling Laws for Thermonuclear Fusion Power Multiplication

While present experiments are evaluated on the basis of confinement time, it is the fusion power multiplication factor, Q, and the fusion power that will be the parameters measuring the performance of ignition experiments and fusion reactors. We have determined the relationship of Q to /tau/ /SUB E/ and the Lawson number, /eta//tau/ /SUB E/ , for ohmically heated plasmas from the Princeton large tokamak (PLT). The values Q, /tau/ /SUB E/ , and /eta//tau/ /SUB E/ all increase with density at low densities. Above /eta/ /SUB e/ approx. = 4 X 10/sup 13/ cm/sup -3/, /tau/ /SUB Ee/ approx. = 30 ms, or /eta/ /SUB e/ /tau/ /SUB Ee/ approx. = 1.2 X 10/sup 12/ cm/sup -3/ s, Q saturates; Q scaling has also been obtained on PLT as a function of toroidal magnetic field, plasma current, and auxiliary heating power.

Enhancement of high-energy ion tails in the presence of combined neutral-beam and ICRF heating of a two-component plasma

An investigation is made of the effect of tuning ICRF heating to neutral-beam-injected ions in a two-component plasma. The corresponding RF-induced velocity diffusion is found to cause a significant enhancement of the high-energy tail of the beam ions. This is also shown to give rise to a subsequent increase in the fusion power multiplication factor, indicating an efficient use of the available heating power.

Active burn control of nearly ignited plasmas

The stabilization of thermal runaway in nearly ignited plasmas by actively controlled auxiliary heating is considered. The use of a variable amount of auxiliary heating for burn control can greatly increase the values of the power multiplication factor Q (fusion power/auxiliary heating power) relative to those values permitted with constant auxiliary heating. A one-dimensional calculation is used to determine the maximum allowable deviation from thermal equilibrium as a function of the equilibrium temperature and Q. The results are applied to tokamak plasmas. The effect of the auxiliary heating deposition profile upon Q is determined. For fixed deviation from thermal equilibrium, significantly higher values of Q can be obtained with central ion heating relative to those obtained with edge ion heating. Central electron heating has about the same order of effectiveness as edge ion heating. For central temperatures of ∼ 25 keV, Q ∼ 25 can be obtained if the deviation in the ion temperature is held below 15%.

A hybrid-ring method of simulating higher powers than are available in waveguides

A method is described whereby a waveguide component can be subjected to an effective power which is greater than that actually available from the magnetron; with a single hybrid ring the effective power is twice that available, and when n such rings are used the power multiplication factor is 2n.