Elsevier

Fusion Engineering and Design

Volume 100, November 2015, Pages 378-405
Fusion Engineering and Design

ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets

https://doi.org/10.1016/j.fusengdes.2015.07.008Get rights and content

Highlights

  • ARC reactor designed to have 500 MW fusion power at 3.3 m major radius.

  • Compact, simplified design allowed by high magnetic fields and jointed magnets.

  • ARC has innovative plasma physics solutions such as inboardside RF launch.

  • High temperature superconductors allow high magnetic fields and jointed magnets.

  • Liquid immersion blanket and jointed magnets greatly simplify tokamak reactor design.

Abstract

The affordable, robust, compact (ARC) reactor is the product of a conceptual design study aimed at reducing the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion Pilot power plant. ARC is a ∼200–250 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has rare earth barium copper oxide (REBCO) superconducting toroidal field coils, which have joints to enable disassembly. This allows the vacuum vessel to be replaced quickly, mitigating first wall survivability concerns, and permits a single device to test many vacuum vessel designs and divertor materials. The design point has a plasma fusion gain of Qp  13.6, yet is fully non-inductive, with a modest bootstrap fraction of only ∼63%. Thus ARC offers a high power gain with relatively large external control of the current profile. This highly attractive combination is enabled by the ∼23 T peak field on coil achievable with newly available REBCO superconductor technology. External current drive is provided by two innovative inboard RF launchers using 25 MW of lower hybrid and 13.6 MW of ion cyclotron fast wave power. The resulting efficient current drive provides a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing fluorine lithium beryllium (FLiBe) molten salt. The liquid blanket is low-risk technology and provides effective neutron moderation and shielding, excellent heat removal, and a tritium breeding ratio  1.1. The large temperature range over which FLiBe is liquid permits an output blanket temperature of 900 K, single phase fluid cooling, and a high efficiency helium Brayton cycle, which allows for net electricity generation when operating ARC as a Pilot power plant.

Introduction

Most fusion reactor designs, such as the ARIES studies [1], [2], [3], [4], assume a large, fixed 1000 MWe output for a power plant. However, large-scale designs make fusion engineering research and development difficult because of the high cost and long construction time of experiments. This paper presents a smaller, less costly, timelier, and lower risk alternative, the 200 MWe ARC reactor. ARC is a conceptual point design of a fusion nuclear science facility/Pilot power plant that demonstrates the advantages of a compact, high-field design utilizing REBCO superconducting magnets and inboard launched lower hybrid current drive (LHCD). The design was carried out as a follow-on to the Vulcan conceptual design; a tokamak for studying plasma–material interaction (PMI) physics that also utilized the demountable REBCO tape and high-field side LHCD [5]. A goal of the ARC design is to minimize the reactor size in order to reduce the plant capital cost. Like Vulcan and several other proposed tokamaks [2], [6], [7], [8], ARC makes use of high-temperature superconductors (HTS), which enables large on-axis magnetic fields and ultimately reduces the size of the reactor. It is important to emphasize that ARC represents one of many possible compact, high-field design configurations. As discussed later in this paper, the modular nature of ARC allows it to change experimental direction and pursue the nuclear materials and vacuum vessel configurations that are determined to be most promising. This enables more innovative and speculative designs because the cost and operational implications of failure are reduced. Indeed a starting design philosophy of ARC is that failure should and will occur as various fusion materials and power exhaust technologies are tried and tested. However, because they can be readily fixed, these failures should not compromise the overall capacity of the device to produce fusing plasmas.

This paper is organized in the following way. Section 2 presents an overview of the ARC design. Section 3 describes the plasma physics basis for the reactor and discusses the current drive system. Section 4 details the design of the magnet system. Section 5 presents the design of the fusion power core, consisting of the tritium breeding/heat exchange blanket and the neutron shield. Section 6 presents a simple costing estimate. Section 7 briefly lists the most vital research and development necessary to enable a design similar to ARC. Lastly, Section 8 provides some concluding remarks.

Section snippets

Design motivation and overview

The ARC reactor is a conceptual tokamak design that can function as both a demonstration fusion power plant for energy generation and a fusion nuclear science facility (FNSF) for integrated materials and component irradiation testing in a D-T neutron field. The starting objective of the ARC study was to determine if a reduced size D-T fusion device (fusion power ≤500 MW) could benefit from the high magnetic field technology offered by recently developed high temperature superconductors. The

0-D point design optimization

In order to determine a starting point for the ARC parameters, a 0-D design exercise was performed. After the initial parameters in this section were determined, the design was iterated several times using codes such as ACCOME, MCNP, and COMSOL. Note that in many cases the final design parameters (e.g. in Table 1 and in the sections following this one) differ from the initial parameters calculated in this section. A fundamental equation for any magnetic fusion reactor design is the scaling [22]

P

Magnet design

A central aspect of the ARC conceptual design is exploring possible fusion reactor/FNSF scenarios at the much higher field afforded by REBCO superconductors. It is imperative to explore these new magnet designs to understand the tradeoffs and limitations. The magnet system, shown in Fig. 17, is divided into four groups: toroidal field (TF) coils, poloidal field (PF) coils, the central solenoid (CS), and auxiliary (AUX) coils. The first two groups are steady state superconducting magnets that

Fusion power core

Traditional tritium breeding and neutron absorbing blankets for fusion reactor designs involve complex components, including significant solid, structural material. Since the blanket is generally contained within the TF coils, these structures must also be separable into toroidal sections so they can be installed through access ports between the TF coils. This results in challenging engineering constraints, difficult remote handling, and a low tritium breeding ratio (TBR) because the structural

Economics

The main driver for minimizing the size of ARC is to reduce the cost of building the reactor. While a full costing of the ARC reactor is beyond the scope of this paper, a rough costing based on volumes and materials prices has been performed. With a major radius of 3.3 m, ARC is similar in size to experiments that have already been built (JET and TFTR). The following analysis aims to justify that ARC is feasible from a materials cost standpoint.

In order to assess the bulk materials costs of the

Plasma physics and current drive

First, the I-mode regime must be further studied, characterized, and demonstrated with non-inductive profiles. As with all small reactor designs the core scenario exploits enhanced confinement from current profile and q control. Therefore, a fully developed and consistent non-inductive scenario with the required physics parameters should be explored more completely. Ideally, we would use a burning plasma experiment in order to also test the self-determining effect of alpha-dominated heating on

Conclusions

With a major radius of 3.3 m and minor radius of 1.1 m, ARC is significantly smaller in size and thermal output than most current reactor designs, which typically generate ∼1 GWe. ARC produces 525 MW of fusion power (∼ 200 MWe), operating in the promising I-mode regime. Steady state plasma current is driven by ICRF fast wave and lower hybrid waves, both launched from the high field side. The reactor has a bootstrap fraction of only 63%, which gives operators greater control of the current profile.

Acknowledgements

We thank Leslie Bromberg, Charles Forsberg, Martin Greenwald, Amanda Hubbard, Brian LaBombard, Bruce Lipschultz, Earl Marmar, Joseph Minervini, Geoff Olynyk, Michael Short, Pete Stahle, Makoto Takayasu, and Stephen Wolfe for conversations and comments that improved this paper. We also thank Zach Hartwig for allowing us to use his C++ wrapper for MCNP and for advice regarding neutronics. BNS was supported by U.S. DoE Grant No. DE-FG02-94ER54235. JB was supported by U.S. DoE Grant No. DE-SC008435

References (115)

  • S. Jardin et al.

    Physics basis for the advanced tokamak fusion power plant, ARIES-AT

    Fusion Eng. Des.

    (2006)
  • H. Shin et al.

    Reversible tensile strain dependence of the critical current in YBCO coated conductor tapes

    Phys. C: Supercond.

    (2007)
  • A. Möeslang et al.

    The IFMIF test facilities design

    Fusion Eng. Des.

    (2006)
  • L. Bromberg et al.

    Options for the use of high temperature superconductor in tokamak fusion reactor designs

    Fusion Eng. Des.

    (2001)
  • L.A. El-Guebaly et al.

    Toward the ultimate goal of tritium self-sufficiency: Technical issues and requirements imposed on ARIES advanced power plants

    Fusion Eng. Des.

    (2009)
  • S. Sato et al.

    Impact of armor materials on tritium breeding ratio in the fusion reactor blanket

    J. Nucl. Mater.

    (2003)
  • S. Delpech et al.

    Molten fluorides for nuclear applications

    Mater. Today

    (2010)
  • D. Whyte et al.

    Reactor similarity for plasma-material interactions in scaled-down tokamaks as the basis for the Vulcan conceptual design

    Fusion Eng. Des.

    (2012)
  • F. Najmabadi

    The ARIES-I tokamak reactor study

    Fusion Technol.

    (1991)
  • B. Coppi et al.

    Critical physics issues for ignition experiments: Ignitor

    MITRLE Report PTP99/06

    (1999)
  • T. Ando et al.

    Design of the tf coil for a tokamak fusion power reactor with ybco tape superconductors

  • D. Kingham, A. Sykes, M. Gryaznevich, Efficient compact fusion reactor, US Patent App. 14/240,809 (Aug. 24,...
  • J. Menard et al.

    Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    Nucl. Fusion

    (2011)
  • R. Stambaugh et al.

    Candidates for a fusion nuclear science facility (FDF and ST-CTF)

  • M. Greenwald et al.

    A new look at density limits in tokamaks

    Nucl. Fusion

    (1988)
  • F. Troyon et al.

    MHD-limits to plasma confinement

    Plasma Phys. Controll. Fusion

    (1984)
  • ITER Physics Basis Editors

    Plasma confinement and transport

    Nucl. Fusion

    (1999)
  • R. Aymar

    Summary of the ITER final design report

    ITER document G A0 FDR

    (2001)
  • R. Crossland et al.

    COMPASS TF coil dynamic vertical preload device and PF coil alignment using a fixed coil array

    Fusion Technol.

    (1990)
  • W. Beck

    Alcator C-MOD toroidal field magnet assembly

  • J. Mattingly

    Elements of Gas Turbine Propulsion, AIAA education series

    (2005)
  • K. Clarno et al.

    Trade studies for the liquid-salt-cooled very high-temperature reactor

    Fiscal year 2006 progress report

    (2007)
  • J. Wesson

    Tokamaks

    (2004)
  • M. Sugihara et al.

    Edge safety factor at the onset of plasma disruption during VDEs in JT-60U

    Plasma Phys. Controll. Fusion

    (2004)
  • R. Stambaugh et al.

    Relation of vertical stability and aspect ratio in tokamaks

    Nucl. Fusion

    (1992)
  • R.J. Thome et al.

    Mhd and fusion magnets: field and force design concepts

    (1982)
  • N. Fisch et al.

    Creating an asymmetric plasma resistivity with waves

    Phys. Rev. Lett.

    (1980)
  • N. Pomphrey

    Bootstrap Dependence on Plasma Profile Parameters, Tech. rep.

    (1992)
  • C. Kessel

    Bootstrap current in a tokamak

    Nucl. Fusion

    (1994)
  • D. Whyte et al.

    the Alcator C-Mod Team, I-mode: an H-mode energy confinement regime with L-mode particle transport in Alcator C-Mod

    Nucl. Fusion

    (2010)
  • A.E. Hubbard et al.

    the Alcator C-Mod Group, Edge energy transport barrier and turbulence in the I-mode regime on Alcator C-Mod

    Phys. Plasmas

    (2011)
  • F. Wagner et al.

    Regime of improved confinement and high beta in neutral-beam-heated divertor discharges of the ASDEX tokamak

    Phys. Rev. Lett.

    (1982)
  • J.R. Walk

    Pedestal structure and stability in high-performance plasmas on Alcator C-Mod

    (2014)
  • M. Bécoulet et al.

    Edge localized mode physics and operational aspects in tokamaks

    Plasma Phys. Controll. Fusion

    (2003)
  • A. Dominguez

    Study of density fluctuations and particle transport at the edge of I-mode plasmas

    (2012)
  • D. Whyte et al.

    I-mode for ITER?

  • P. Stangeby

    The plasma boundary of magnetic fusion devices

  • T. Taylor et al.

    Profile optimization and high beta discharges, and stability of high elongation plasmas in the DIII-D tokamak

  • A.E. Hubbard et al.

    Threshold conditions for transitions to I-Mode and H-Mode with unfavorable ion grad B drift direction

    Nucl. Fusion

    (2012)
  • J. Walk et al.

    Edge-localized mode avoidance and pedestal structure in I-mode plasmas

    Phys. Plasmas (1994-present)

    (2014)
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