Fusion Ignition Research Experiment Research Articles

Smaller & Sooner: Exploiting High Magnetic Fields from New Superconductors for a More Attractive Fusion Energy Development Path

The current fusion energy development path, based on large volume moderate magnetic B field devices is proving to be slow and expensive. A modest development effort in exploiting new superconductor magnet technology development, and accompanying plasma physics research at high-B, could open up a viable and attractive path for fusion energy development. This path would feature smaller volume, fusion capable devices that could be built more quickly than low-to-moderate field designs based on conventional superconductors. Fusion’s worldwide development could be accelerated by using several small, flexible devices rather than relying solely on a single, very large device. These would be used to obtain the acknowledged science and technology knowledge necessary for fusion energy beyond achievement of high gain. Such a scenario would also permit the testing of multiple confinement configurations while distributing technical and scientific risk among smaller devices. Higher field and small size also allows operation away from well-known operational limits for plasma pressure, density and current. The advantages of this path have been long recognized—earlier US plans for burning plasma experiments (compact ignition tokamak, burning plasma experiment, fusion ignition research experiment) featured compact high-field designs, but these were necessarily pulsed due to the use of copper coils. Underpinning this new approach is the recent industrial maturity of high-temperature, high-field superconductor tapes that would offer a truly “game changing” opportunity for magnetic fusion when developed into large-scale coils. The superconductor tape form and higher operating temperatures also open up the possibility of demountable superconducting magnets in a fusion system, providing a modularity that vastly improves simplicity in the construction, maintenance, and upgrade of the coils and the internal nuclear engineering components required for fusion’s development. Our conclusion is that while tradeoffs exist in design choices, for example coil, cost and stress limits versus size, the potential physics and technology advantages of high-field superconductors are attractive and they should be vigorously pursued for magnetic fusion’s development.

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5 mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by the commercial finite element code, COSMOS. The heat transfer coefficients at the nodes located on the cooling channel-wall are determined outside COSMOS code by the same methodology used in ITERTHA. A good agreement is achieved under different incident heat fluxes. The ITERTHA code is also benchmarked against the thermal-hydraulic calculation of the outer divertor of the Fusion Ignition Research Experiment, FIRE for an incident heat flux of 20 MW/m 2 and coolant flow velocity of 10 m/s in a cooling channel of 8 mm diameter with swirl-tape inserts of 2 ratio and 1.5 mm thickness. The results show excellent agreement for both steady and transient states and prove the successful implementation of both the hydraulic and heated diameters of the swirl-tape channels in the used heat transfer correlations.

Erosion, transport, and tritium codeposition analysis of a beryllium wall tokamak

We analyzed beryllium first wall sputtering erosion, sputtered material transport, and T/Be codeposition for a typical next-generation tokamak design—the fusion ignition research experiment (FIRE). The results should be broadly applicable to any future tokamak with a beryllium first wall. Starting with a fluid code scrapeoff layer attached plasma solution, plasma D0 neutral fluxes to the wall and divertor are obtained from the DEGAS2 neutral transport code. The D+ ion flux to the wall is computed using both a diffusive term and a simple convective transport model. Sputtering coefficients for the beryllium wall are given by the VFTRIM-3D binary-collision code. Transport of beryllium to the divertor, plasma, and back to the wall is calculated with the WBC+ code, which tracks sputtered atom ionization and subsequent ion transport along the SOL magnetic field lines. Then, using results from a study of Be/W mixing/sputtering on the divertor, and using REDEP/WBC impurity transport code results, we estimate the divertor surface response. Finally, we compute tritium codeposition rates in Be growth regions on the wall and divertor for D–T plasma shots using surface temperature dependent D–T/Be rates and with different assumed oxygen contents. Key results are: (1) peak wall net erosion rates vary from about 0.3 nm s−1 for diffusion-only transport to 3 nm s−1 for diffusion plus convection, (2) T/Be codeposition rates vary from about 0.1 to 10.0 mg T s−1 depending on the model, and (3) core plasma contamination from wall-sputtered beryllium is low in all cases (< 0.02%). Thus, based on the erosion and codeposition results, the performance of a beryllium first wall is very dependent on the plasma response, and varies from acceptable to unacceptable.

Burning plasma projections using drift-wave transport models and scalings for the H-mode pedestal

The GLF23 and multi-mode core transport models are used along with models for the H-mode pedestal to predict the fusion performance for the International Thermonuclear Experimental Reactor, Fusion Ignition Research Experiment, and IGNITOR tokamak designs. Simulations using combinations of core and pedestal models have also been compared with experimental data for H-mode profiles in DIII-D, JET, and Alcator C-Mod. Power-independent (ballooning mode limit) and power-dependent pedestal scalings lead to very different predictions when used with the core models. Although the two drift-wave transport models reproduce the core profiles in a wide variety of tokamak discharges, they differ in their projections to burning plasma experiments for the same pedestal parameters. Differences in the core transport models in their response to the ion temperature gradient (i.e. their stiffness) and impact of the power dependence of the H-mode pedestal on fusion performance predictions are discussed.

Safety in the Design of Three Burning Plasma Experiments

The 2002 Snowmass Fusion Energy Sciences Summer Study required a uniform assessment of the safety design goals for three candidate burning plasma experiments: the Fusion Ignition Research Experiment (FIRE), the IGNITOR compact tokamak, and the International Thermonuclear Experimental Reactor (ITER). The main assessment criterion was an objective judgment of each design’s ability to obtain a generalized regulatory approval. A brief overview of environmental impact, safety, and health results from the uniform assessment of safety are given in this paper. As safety documentation was reviewed for each design, several issues became apparent. This paper also documents these specific issues. Each of these three designs could obtain a general regulatory approval based on their safety design practices.

Study of thermonuclear Alfvén instabilities in next step burning plasma proposals

The stability of α-particle driven shear Alfvén eigenmodes (AE) for nominal burning plasma (BP) parameters in the proposed international tokamak experimental reactor (ITER), fusion ignition research experiment (FIRE) and IGNITOR tokamaks is studied. JET plasma, where fusion αs were generated in tritium experiments, is also studied to compare the numerical predictions with the existing experiments. An analytic assessment of toroidal AE (TAE) stability is first presented, where the α-particle β due to the fusion reaction rate and electron drag is simply and accurately estimated in plasmas with central temperature in the range of 7–20 keV. In this assessment the hot particle drive is balanced against ion-Landau damping of the background deuterons, and electron collision effects and stability boundaries are determined. Then two numerical studies of AE instability are presented. In one, the HIgh-n STability (HINST) code is used to predict the instabilities of low and moderately high frequency Alfvén modes. HINST computes the non-perturbative solutions of the AE including effects of ion finite Larmor radius, orbit width, trapped electrons etc. The stability calculations are repeated using the global code NOVAK. We show that for these plasmas the spectrum of the least stable AE modes is at medium-/high-n numbers. In HINST, TAEs are locally unstable due to the α pressure gradient in all the devices under consideration except IGNITOR. However, NOVAK calculations show that the global mode structure enhances the damping mechanisms and produces stability for the nominal FIRE proposal and near-marginal stability for the nominal ITER proposal. NBI ions produce a strong stabilizing effect for JET. However, in ITER, the beam energies needed to penetrate to the core must be high (∼1 MeV) so that a diamagnetic drift frequency comparable to that of α-particles is produced by the beam ions which induces a destabilizing effect. A serious question remains whether the perturbation theory used in NOVAK overestimates the stability predictions, so that it is premature to conclude that the nominal operation of all three BP proposals without neutral beam injection are stable (or marginally stable) to AEs.

Development of novel radiation-resistant insulation systems for fusion magnets

The development of higher performance composite insulation systems for use in fusion magnets has been an important goal for the fusion community in recent years. Next Step Option (NSO) fusion devices, such as the Fusion Ignition Research Experiment (FIRE), are being designed with the assumption that new, better performing insulation systems will be available at the time of magnet manufacture. To address these concerns, Composite Technology Development, Inc. (CTD) has developed a new class of organic composite insulation systems designed not only to meet the performance criteria of these new magnets, but also to meet the fabrication challenges that will be encountered during magnet fabrication. These new systems, based on previous work with cyanate ester resin systems, have been developed with a focus on increased radiation-resistance, ease of processing and fabrication, and mechanical and electrical strength at cryogenic and elevated temperatures. New resin systems have been developed to enable a broad range of insulation application methods including vacuum pressure impregnation (VPI) and pre-impregnation. Processing information on these systems, along with their mechanical and electrical properties will be presented.

Engineering overview of the fusion ignition research experiment (FIRE) : Thome, R. J. et al. Fusion Engineering and Design, 2002, 63–64, 53–57

Engineering overview of the fusion ignition research experiment (FIRE) : Thome, R. J. et al. Fusion Engineering and Design, 2002, 63–64, 53–57

Plasma heating and current drive systems for the fusion ignition research experiment (FIRE) : Swain, D. W. and Carter, M. D. Fusion Engineering and Design, 2002, 63–64, 541–545

Plasma heating and current drive systems for the fusion ignition research experiment (FIRE) : Swain, D. W. and Carter, M. D. Fusion Engineering and Design, 2002, 63–64, 541–545

Diagnostics for FIRE: A status report

The mission for the proposed fusion ignition research experiment (FIRE) device is to “attain, explore, understand and optimize fusion-dominated plasmas.” Operation at Q⩾5, for 20 s with a fusion power output of ∼150 MW is the major goal. Attaining this mission sets demands for plasma measurement which are at least as comprehensive as on present tokamaks, with the additional capabilities needed for control of the plasma and for understanding the effects of the alpha-particles. Because of the planned operation in advanced tokamak scenarios, with steep transport barriers, the diagnostic instrumentation must be able to provide fine spatial and temporal resolution. It must also be able to withstand the impact of the intense neutron and gamma irradiation. There are practical engineering issues of minimizing radiation streaming while providing essential diagnostic access to the plasma. Many components will operate close to the first wall, e.g., ceramics and mineral insulated cable for magnetic diagnostics and mirrors for optical diagnostics; these components must be selected and mounted so that they will operate and survive in fluxes which require special material selection. The measurement requirements have been assessed so that the diagnostics for the FIRE device can be defined. Clearly, a better set of diagnostics of alpha-particles than that available for TFTR is essential, since the alpha-particles provide the dominant sources of heating and of instability-drive in the plasma.

FIRE, a next step option for magnetic fusion

The next major frontier in magnetic fusion physics is to explore and understand the strong non-linear coupling among confinement, MHD stability, self-heating, edge physics and wave-particle interactions that is fundamental to fusion plasma behavior. The Fusion Ignition Research Experiment (FIRE) design study has been undertaken to define the lowest cost facility to attain, explore, understand and optimize magnetically confined fusion-dominated plasmas. FIRE is envisioned as an extension of the existing advanced tokamak (AT) program that could lead to an attractive magnetic fusion reactor. FIRE activities have focused on the physics and engineering assessment of a compact, high-field tokamak with the capability of achieving Q≈10 in the Elmy H-mode for a duration of ∼1.5 plasma current redistribution times (skin times) during an initial burning plasma science phase, and the flexibility to add AT hardware (e.g. lower hybrid current drive) later. The configuration chosen for FIRE is similar to that of ARIES-RS, the US Fusion Power Plant study utilizing an AT reactor. The key ‘AT’ features are: strong plasma shaping, double null pumping divertors, low toroidal field (TF) ripple (<0.3%), internal control coils and space for wall stabilization capabilities. The reference design point is R o=2.14 m, a=0.595 m, B t( R o)=10 T, I p=7.7 MA with a flat top time of 20 s for 150 MW of fusion power. The baseline magnetic fields and pulse lengths can be provided by wedged BeCu/OFHC TF coils and OFHC poloidal field (PF) coils that are pre-cooled to 80 K prior to the pulse and allowed to warm up to 373 K at the end of the pulse. A longer term goal of FIRE is to explore AT regimes sustained by non-inductive current drive (e.g. lower hybrid current drive) at high fusion gain ( Q>5) for a duration of one to three current redistribution times.

Nuclear features of the fusion ignition research experiment (FIRE)

The main nuclear features of the baseline design of fusion ignition research experiment (FIRE) have been evaluated. Critical issues were addressed and R&D needs were identified. Modest values of nuclear heating occur in the FIRE components. The total nuclear heating in the 16 TF coils during DT shots is 19 MW. The cumulative damage in the copper alloys used is very low (<0.05 dpa). However, issues of low temperature embrittlement and thermal creep at high temperatures need to be resolved by an R&D program. The radiation induced resistivity increase in the conductors of the TF coils is primarily due to displacement damage and is <20% of the unirradiated resistivity. The magnet insulator development R&D program should involve irradiation to dose levels up to 1.5×10 10 Rad with the proper mix between neutrons and gamma photons (50% gamma dose) relevant to FIRE conditions. Low levels of activity and decay heat are obtained. Hands-on ex-vessel maintenance is feasible. All components qualify as Class C low-level waste. Activation of nitrogen gas inside the cryostat produces a very small amount of 13N and 14C.

Physics basis and simulation of burning plasma physics for the fusion ignition research experiment (FIRE)

The FIRE design for a burning plasma experiment is described in terms of its physics basis and engineering features. Systems analysis indicates that the device has a wide operating space to accomplish its mission, both for the ELMy H-mode reference and the high bootstrap current/high-β advanced tokamak regimes. Simulations with 1.5D transport codes reported here both confirm and constrain the systems projections. Experimental and theoretical results are used to establish the basis for successful burning plasma experiments in FIRE.

Engineering overview of the fusion ignition research experiment (FIRE)

The FIRE tokamak is an option for the next step in the U.S. magnetic fusion energy program. The design goals have evolved to a major radius of 2.14 m, minor radius of 0.525 m, toroidal field (TF) of 10 T and plasma current of 7.7 mA for a flat-top time of ∼20 s. The requirement for 3000 full-power, full-field pulses and 30,000 pulses at 2/3 of full field has been retained since 1999. All magnets are inertially cooled with liquid nitrogen. Design options have been considered over a range of parameter space for either a wedged TF coil structure or a bucked and wedged TF/Central Solenoid structure. The wedged configuration has been selected as the baseline.

Plasma heating and current drive systems for the fusion ignition research experiment (FIRE)

FIRE is a high-field, burning-plasma tokamak that is being studied as a possible option for future fusion research. Preliminary parameters for this machine are R 0=2.14 m, a=0.595 m, B 0=10 T, and I p=7.7 MA. Magnetic field coils are to be made of copper and pre-cooled with LN 2 before each shot. The flat-top pulse length desired is ≥20 s. Ion cyclotron (IC) and lower hybrid rf systems will be used for heating and current drive. Present specifications call for 20 MW of IC heating power, with an additional 10 MW of IC power or 20 MW of lower hybrid power as upgrade options.

Prediction of density limits in tokamaks: Theory, comparison with experiment, and application to the proposed Fusion Ignition Research Experiment

A framework for the predictive calculation of density limits in future tokamaks is proposed. Theoretical models for different density limit phenomena are summarized, and the requirements for additional models are identified. These theoretical density limit models have been incorporated into a relatively simple, but phenomenologically comprehensive, integrated numerical calculation of the core, edge, and divertor plasmas and of the recycling neutrals, in order to obtain plasma parameters needed for the evaluation of the theoretical models. A comparison of these theoretical predictions with observed density limits in current experiments is summarized. A model for the calculation of edge pedestal parameters, which is needed in order to apply the density limit predictions to future tokamaks, is summarized. An application to predict the proximity to density limits and the edge pedestal parameters of the proposed Fusion Ignition Research Experiment is described.

Engineering features of the fusion ignition research experiment (FIRE)

The fusion ignition research experiment (FIRE) tokamak is an option for the next step in the US magnetic fusion energy program. It has a major radius of 2 m, and a minor radius of 0.525 m. The general requirements for FIRE are B T=10 T; I P=6.5 MA; minimum flat top time=10 s, and maximum number of full-power, full-field pulses=3000 with 30 000 pulses at 2/3 full field. The baseline design is able to meet or exceed these objectives. All magnets are inertially cooled with liquid nitrogen and have the capability for an 18 s flat top at 10 T. The use of BeCu in the inner legs of the toroidal field (TF) coil also allows for a field of 12 T with a pulse length of 12 s. Extended pulse lengths at lower fields (e.g. 214 s at 4 T and 2 MA) will allow FIRE to explore advanced tokamak modes (Physics basis for the Fusion Ignition Research Experiment (FIRE) Plasma Facing Components, in: Twenty-first Symposium on Fusion Technology, Madrid, September, 2000).

Physics basis for the fusion ignition research experiment plasma facing components

A design study of a fusion ignition research experiment (FIRE) is underway to investigate and assess near term opportunities for advancing the scientific understanding of self-heated fusion plasmas. The emphasis for the FIRE program is on understanding the behavior of plasmas dominated by alpha heating ( Q≥5). Study activities have focused on the technical evaluation of a compact, high field, highly shaped tokamak. One of the key issues for the design is to find suitable plasma facing components (PFCs). We have investigated a variety of plasma edge and divertor conditions ranging from reduced recycling high heat flux conditions (attached) to reduced heat flux detached operation. The inner divertor detaches easily while impurities must be added to the outer divertor to achieve detachment. The outer divertor and private space baffle will have to be actively cooled. The plasma-facing surface of the divertor is tungsten bonded to a CuCrZr heat sink. The remainder of the PFCs are beryllium coated copper attached to the vacuum vessel. Plasma current disruptions impose strong constraints on the design. Appreciable PFC surface melting and evaporation and onset of ‘plasma shielding’ are expected. The forces induced on the PFC due to disruptions determine the size of the attachment of the PFC to the vacuum vessel.

FIRE Facilities and Site Requirements

This paper describes the buildings and balance of plant systems required to support the Fusion Ignition Research Experiment (FIRE) Project. Facilities and systems are developed on the basis of a “greenfield” site, with no benefit for existing facilities, but also without any constraints on the potential arrangement. Because FIRE will operate deuterium-tritium plasmas for pulse lengths on the order of 20 seconds, FIRE will require a moderate on-site tritium inventory. FIRE buildings and systems must be designed and licensed to comply with regulations for nuclear facilities. They must also include systems to manage tritium and tritiated water, activated dust, and radioactive waste material. Maintenance activities on FIRE will require the use of remote handling systems to remove and transport tokamak parts to hot cell facilities. Major tokamak service connections will be required to feed power to the copper magnet system and deliver plasma-heating energy to ICRF antennae. Competition for access to the tokamak for service connections and repair activities will constrain the overall arrangement and routing of services.This paper examines the design implications for the fuel supply, vacuum pumping, fuel recovery, cooling, and other balance of plant systems that contribute to the control of radioactive materials. It also examines the design implications for the tokamak test cell, hot cells, structures to house key services, and routing of service connections to the tokamak. Site requirements, a generic site plan, and conceptual building arrangements are provided.

Thermal Hydraulic Analysis of FIRE Divertor

The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the U.S. magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating.The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m2. A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush.The finite element (FE) analysis shows that for an inlet water temperature of 30°C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m2. The peak copper temperature is 490°C, peak tungsten temperature is 1560°C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements.