Fusion Reactor Design Studies Research Articles

GX: a GPU-native gyrokinetic turbulence code for tokamak and stellarator design

GX is a code designed to solve the nonlinear gyrokinetic system for low-frequency turbulence in magnetized plasmas, particularly tokamaks and stellarators. In GX, our primary motivation and target is a fast gyrokinetic solver that can be used for fusion reactor design and optimization along with wide-ranging physics exploration. This has led to several code and algorithm design decisions, specifically chosen to prioritize time to solution. First, we have used a discretization algorithm that is pseudospectral in the entire phase space, including a Laguerre–Hermite pseudospectral formulation of velocity space, which allows for smooth interpolation between coarse gyrofluid-like resolutions and finer conventional gyrokinetic resolutions and efficient evaluation of a model collision operator. Additionally, we have built GX to natively target graphics processors (GPUs), which are among the fastest computational platforms available today. Finally, we have taken advantage of the reactor-relevant limit of small $\rho _*$ by using the radially local flux-tube approach. In this paper we present details about the gyrokinetic system and the numerical algorithms used in GX to solve the system. We then present several numerical benchmarks against established gyrokinetic codes in both tokamak and stellarator magnetic geometries to verify that GX correctly simulates gyrokinetic turbulence in the small $\rho _*$ limit. Moreover, we show that the convergence properties of the Laguerre–Hermite spectral velocity formulation are quite favourable for nonlinear problems of interest. Coupled with GPU acceleration, which we also investigate with scaling studies, this enables GX to be able to produce useful turbulence simulations in minutes on one (or a few) GPUs and higher fidelity results in a few hours using several GPUs. GX is open-source software that is ready for fusion reactor design studies.

Recent development and application of a new safety analysis code for fusion reactors

This paper describes the recent progress made in the development of two codes for fusion reactor safety assessments at the Idaho National Laboratory (INL): MELCOR for fusion and the Tritium Migration Analysis Program (TMAP). During the ITER engineering design activity (EDA), the INL Fusion Safety Program (FSP) modified the MELCOR 1.8.2 code for fusion applications to perform ITER thermal hydraulic safety analyses. Because MELCOR has undergone many improvements at SNL-NM since version 1.8.2 was released, the INL FSP recently imported these same fusion modifications into the MELCOR 1.8.6 code, along with the multiple fluids modifications of MELCOR 1.8.5 for fusion used in US advanced fusion reactor design studies. TMAP has also been under development for several decades at the INL by the FSP. TMAP treats multi-specie surface absorption and diffusion in composite materials with dislocation traps, plus the movement of these species from room to room by fluid flow within a given facility. Recently, TMAP was updated to consider multiple trap site types to allow the simulation of experimental data from neutron irradiated tungsten. The natural development path for both of these codes is to merge their capabilities into one computer code to provide a more comprehensive safety tool for analyzing accidents in fusion reactors. In this paper we detail recent developments in this area and the application of this new capability to a DEMO relevant water-cooled tungsten armored divertor.

Tokamak plasma edge modelling including the main chamber wall

Quantifying main chamber wall recycling, erosion and resulting material migration, at least on the basis of known or empirical far scrape-off layer (SOL) processes, is still highly uncertain, despite its relevance for ITER and fusion reactor design studies. This affects, for example, the design problem of first mirror performance of many optical diagnostics in the harsh ITER environment. Poor computational access is not least due to a fundamental technical limitation in apparently all current tokamak edge plasma fluid codes, which implicates a wide computationally unresolved gap between the outermost plasma layer treated in codes and the real vessel wall. We show how the current ITER version of the B2-EIRENE code (SOLPS-4.3) can be extended to cover also this far SOL, on the same footing as the rest of the plasma transport model. We discuss consequences of this new model for estimating plasma power and particle sink terms caused by a fairly realistic wall in ITER based on the conventional Bohm criterion along all plasma–wall interfaces.Corrections were made to this article on 14 July 2011. The authors have been assigned to the correct affiliations.

Neutronic investigation on the ARIES-ST fusion reactor with fissionable molten salts

There have been a number of major fusion reactor design studies especially during the last 35 years. One of them is ARIES-ST fusion reactor, a 1000 MW el fusion reactor power plant [1–10]. Its design is based on the spherical tokamak concept [1–10]. Furthermore, it has an advantage of having compact power core [1–10]. In ARIES-ST fusion reactor, the coolant of LiPb is considered to be used as both energy carrier and tritium breeder [1–10]. The current study investigates the main neutronic parameters of the ARIES-ST fusion reactor utilizing the molten salts with fissile isotopes; Flibe + Weapon Grade (WG) PuF 4 or Flibe + Spent Fuel Grade (SFG) PuF 4. The neutronic calculations in the fusion blanket were carried out with the help of Scale 5 for various coolant compositions. Numerical results showed that the best neutronic performance in the reactor was attained with the coolant of Flibe + 1% WG PuF 4. In addition, a significant improvement in the energy multiplication of the reactor was obtained in comparison to the pure ARIES-ST fusion reactor.

Overview of the Japanese fusion reactor studies programme

Fusion reactor design studies conducted in Japan since the fifth IAEA Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology in 1993 are introduced. During the 5 years since the last IAEA-TCM, the fusion reactor studies in Japan became more focused concentrating on tokamaks, helical devices and laser fusion. There are four tokamak reactor studies, namely, A-SSTR, CREST, DREAM and IDLT. There are helical reactor studies such as FFHR and one laser fusion reactor study on KOYO. Compared with the situation of 5 years ago having many types of confinement schemes being studied, the confinement schemes are now limited to three and each of the effort is larger involving greater number of scientists. The cooperation and exchanges between the design groups have increased through various committees, meetings and symposia.

An assessment of the prospects for economic D-T tokamak fusion

The purpose of the fusion programme is to provide broadly competitive electricity (with perhaps credit given for possible environmental, diversity and safety benefits). Fusion reactor design studies routinely predict electricity prices broadly comparable to conventional alternatives such as the PWR, despite the devices producing the electricity being typically ∼ 10 times bigger, more complicated and requiring more maintenance. In this paper this issue is considered. These low power densities, indicated both by reactor studies and confirmed by generic engineering and geometric arguments, are shown to be a rather fundamental feature of the D-T tokamak. Coupled with the complex construction and expensive materials required, and the probable low availability its serial complexity and high maintenance needs will cause, it is argued that D-T tokamak electricity prices will be a significant multiple of alternatives. This is the case even with all remaining physics issues assumed resolved successfully.

Ceramic pebble bed development for fusion blankets

Research on lithium ceramic breeders has been intensive since the late 1970s. The bulk material properties of several candidate lithium ceramics are essentially available, although there is still much work to be done on properties under irradiation and on the overall behavior in blanket modules. Based on these results, lithium ceramic breeder blankets have been selected in many fusion reactor design studies. These lithium ceramics are incorporated into blankets typically as monolithic pellets or packed pebble beds. There is substantial industrial experience with pebble beds made from other ceramics as catalyst supports, and in fabrication and testing of pebbles for advanced fission reactor fuels. In fusion blankets, the pebble bed form offers several attractive features, including simpler assembly into complex geometries, a uniform pore network and low sensitivity to cracking or irradiation damage. Ceramic breeder pebbles have been a focus for several research groups. In general, the database is similar to that of monolithic pellets for the materials studied; basic production and material property data are available, but the irradiation and engineering database remains sparse. In addition to the basic requirements on any ceramic breeder material (such as low tritium hold-up, compatibility with structure, irradiation stability, etc.), the main pebble bed requirements may be roughly summarized as follows: economic, high yield production rates; high average bed (smear) density; adequate bed thermal conductivity; acceptable purge gas pressure drop; adequate crush strength; tolerance to thermal cycling. In this paper, the international ceramic breeder pebble bed database is reviewed with respect to these pebble bed properties, and the R&D needs for reactor blanket development are assessed.

Fusion reactor design studies

Abstract This paper is a summary of fusion reactor design studies presented and discussed at the IAEA Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology, held in September 1993 at the University of California, Los Angeles. This summary does not represent all the reactor design studies carried out in the world since the last meeting in Yalta in 1986. The reactor designs presented are mostly D–T fuel cycle with both magnetic confinement and inertial confinement. As a D–3He, only the ARIES-III was presented. Steady state and pulsed tokamak studies, stellarator and reversed field pinch reactor studies are summarized. Six recent D–T inertial confinement fusion reactors presented at the meeting are summarized. A short summary on D–3He reactors and fusion-fission hybrid reactor designs is given.

Integrated fusion environmental and safety systems studies

An integrated environmental and safety (E&S) systems module is being developed at the Berkeley Fusion Environmental and Safety Group with the objectives of providing overall E&S assessment and incorporating E&S characteristics into optimization in fusion reactor design studies. The module includes features such as modularization, quantification of figures-of-merit, and fast calculational schemes because of emphasis on compatibility with systems-level analyses. Initial results indicate the potential for trade-offs and optimizations involving E&S and reactor performance indices in moderately large design windows.

Magnetic Fusion for Space Propulsion

Magnetic fusion could enable the efficient, large-scale exploration and development of the Solar System. Several conceptual fusion reactor design studies indicate that magnetic fusion may be attrac...

Engineering design conditions and constraints affecting experimental fusion reactors and power-reactors

In the design of advanced technology products such as fusion reactors it is important to make clear the criteria which depend on engineering conditions and constraints. For fusion reactor design studies such as ITER, the conditions and constraints for superconducting magnets, high heat flux materials, and so on, have been examined. Using the present database, these issues are reviewed and major considerations for fusion power reactors are discussed. The effect of improving the engineering design constraints is studied by the tokamak system analysis code. It is shown that the development of high heat flux materials, high field magnets and special materials for magnets is the key to the future of fusion reactors.

Advanced Fusion Fuel Cycles

The discovery of the favorable energy balance for {sup 3}He mining on the Moon has led to a profound change in fusion reactor design studies. The replacement of tritium by {sup 3}He in the plasma has been investigated. This paper discusses how the reduced neutron flux in a D-{sup 3}He reactor has substantial engineering, safety, and environmental advantages over a deuterium-tritium (D-T) reactor. Disadvantages are also addressed.

Conceptual designs of tokamak fusion reactors.

Tokamak fusion reactor design studies conducted in JAERI with support from national laboratories, universities and industries in the fifteen years since 1973 to the present are described. These studies gave considerable impact on the national and international fusion programs. From the proposal of He-cooled Li20 blanket for fusion power reactor in 1973 to the most recent proposal of easily replaceable guard limiter concept to the ITER project, a number of unique proposals that have influenced the world wide fusion reactor design studies are introduced. They are described in the context of major fusion reactor design efforts and design methodology developments of the past fifteen years.

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What have fusion reactor studies done for you today?

The University of Wisconsin examines the fusion program and puts into perspective what return is being made on investments in fusion reactor studies. Illustations show financial support for fusion research from the four major programs, FY'82 expenditures on fusion research, and the total expenditures on fusion research since 1951. Topics discussed include the estimated number of scientists conducting fusion research, the conceptual design study of a fusion reactor, scoping study of a reactor, the chronology of fusion reactor design studies, published fusion reactor studies 1967-1983, conceptual fusion reactor design studies, STARFIRE reference design, MARS central cell, HYLIFE reaction chamber, and selected contributions of reactor design studies to base programs.

Engineering Aspects of Thermonuclear Fusion Reactors

Click to increase image sizeClick to decrease image size Additional informationNotes on contributorsKenneth R. SchultzKenneth R. Schultz is manager of the Fusion Development and Technology Branch at GA Technologies in San Diego, California. He received BS and MS degrees in mechanical engineering from Stanford University and a PhD in nuclear engineering from the University of Florida in 1971. He has worked on the core physics design of high-temperature gas-cooled reactors and since 1975 has been involved with fusion reactor design studies and fusion technology experiments. He is the 1982/1983 chairman of the ANS Fusion Energy Division.

Goals of Fusion Reactor Design Studies

Goals of Fusion Reactor Design Studies

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