Tritium Breeding in Fusion Reactors

Abdou, Mohamed

doi:10.1007/978-94-009-7099-1_63

Cited by 18 publications

(4 citation statements)

References 18 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The design of the T breeder is vital to achieving TBR > 1. The best candidate reactions to produce T are between the D-T neutron from the fusion reaction and lithium, either as Li-6 or (the more commonly occurring) Li-7 [27]. Breeder candidate materials exist as liquids (favoured by ICF schemes without the added complexity of magnetic fields) or as solids, both likely requiring neutron multipliers to increase the number of neutrons for breeding.…”

Section: Shared Technical Challengesmentioning

confidence: 99%

An overview of shared technical challenges for magnetic and inertial fusion power plant development

Chapman

Walkden

2020

Phil. Trans. R. Soc. A.

View full text Add to dashboard Cite

Fusion energy is an area of active development and innovation worldwide, with many design concepts studied, each exhibiting a range of technical challenges. A significant portion of technical challenges will be unique for a given design concept; however, there are several overarching challenges that any design must address to some degree. These include tritium handling and the tritium cycle; materials and their survivability in the high-energy neutron environment of D-T fusion; neutronics and the validation of nuclear data; remote handling and maintenance activities; and integrated holistic approaches to fusion plant design. This paper provides an overview of these aspects for magnetic and inertial fusion approaches with a view to highlighting commonality and the benefits of shared knowledge that this may bring. This article is part of a discussion meeting issue ‘Prospects for high gain inertial fusion energy (part 2)’.

show abstract

Section: Shared Technical Challengesmentioning

confidence: 99%

An overview of shared technical challenges for magnetic and inertial fusion power plant development

Chapman

Walkden

2020

Phil. Trans. R. Soc. A.

View full text Add to dashboard Cite

show abstract

“…In principle, all that is needed is an average tritium-breeding ratio (TBR) greater than 1, but more may be needed when fusion is first introduced into the economy to provide a surplus of tritium to fuel the startup of new reactors. 907 In continuous operation, an average TBR of 1.05 should suffice. From the discussion above, a breeder can readily have a local TBR $ 1.5 (Ref.…”

Section: Blanketmentioning

confidence: 99%

Direct-drive inertial confinement fusion: A review

et al. 2015

View full text Add to dashboard Cite

The direct-drive, laser-based approach to inertial confinement fusion (ICF) is reviewed from its inception following the demonstration of the first laser to its implementation on the present generation of high-power lasers. The review focuses on the evolution of scientific understanding gained from target-physics experiments in many areas, identifying problems that were demonstrated and the solutions implemented. The review starts with the basic understanding of laser–plasma interactions that was obtained before the declassification of laser-induced compression in the early 1970s and continues with the compression experiments using infrared lasers in the late 1970s that produced thermonuclear neutrons. The problem of suprathermal electrons and the target preheat that they caused, associated with the infrared laser wavelength, led to lasers being built after 1980 to operate at shorter wavelengths, especially 0.35 μm—the third harmonic of the Nd:glass laser—and 0.248 μm (the KrF gas laser). The main physics areas relevant to direct drive are reviewed. The primary absorption mechanism at short wavelengths is classical inverse bremsstrahlung. Nonuniformities imprinted on the target by laser irradiation have been addressed by the development of a number of beam-smoothing techniques and imprint-mitigation strategies. The effects of hydrodynamic instabilities are mitigated by a combination of imprint reduction and target designs that minimize the instability growth rates. Several coronal plasma physics processes are reviewed. The two-plasmon–decay instability, stimulated Brillouin scattering (together with cross-beam energy transfer), and (possibly) stimulated Raman scattering are identified as potential concerns, placing constraints on the laser intensities used in target designs, while other processes (self-focusing and filamentation, the parametric decay instability, and magnetic fields), once considered important, are now of lesser concern for mainline direct-drive target concepts. Filamentation is largely suppressed by beam smoothing. Thermal transport modeling, important to the interpretation of experiments and to target design, has been found to be nonlocal in nature. Advances in shock timing and equation-of-state measurements relevant to direct-drive ICF are reported. Room-temperature implosions have provided an increased understanding of the importance of stability and uniformity. The evolution of cryogenic implosion capabilities, leading to an extensive series carried out on the 60-beam OMEGA laser [Boehly et al., Opt. Commun. 133, 495 (1997)], is reviewed together with major advances in cryogenic target formation. A polar-drive concept has been developed that will enable direct-drive–ignition experiments to be performed on the National Ignition Facility [Haynam et al., Appl. Opt. 46(16), 3276 (2007)]. The advantages offered by the alternative approaches of fast ignition and shock ignition and the issues associated with these concepts are described. The lessons learned from target-physics and implosion experiments are taken into account in ignition and high-gain target designs for laser wavelengths of 1/3 μm and 1/4 μm. Substantial advances in direct-drive inertial fusion reactor concepts are reviewed. Overall, the progress in scientific understanding over the past five decades has been enormous, to the point that inertial fusion energy using direct drive shows significant promise as a future environmentally attractive energy source.

show abstract

“…The functions of blankets are tritium breeding, heat removal and radiation shielding. Tritium breeding maintains the supply of fuel in the reactor and deposited heat removal through coolant is converted into electricity [5][6][7]. The functions make the blanket a prime component of fusion reactor sustainability.…”

Section: Introductionmentioning

confidence: 99%

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

Swami¹,

Sharma²,

Danani³

et al. 2022

Plasma Sci. Technol.

View full text Add to dashboard Cite

India has proposed the Helium Cooled Solid Breeder blanket concept as a tritium breeding module for testing in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design of the module is prepared to analyze in more detail. The neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in sub-modules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile in the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg for an ITER Full day operation considering the 400 seconds ON time and 1400 seconds dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the TBM is huge (around 9 GJ) for the entire day operation of the ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, it is will be around 1655 MWh annually. The evaluation is done using the Monte Carlo Radiation transport code and ENDF-VIII nuclear cross-section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.

show abstract

Tritium Breeding in Fusion Reactors

Cited by 18 publications

References 18 publications

An overview of shared technical challenges for magnetic and inertial fusion power plant development

An overview of shared technical challenges for magnetic and inertial fusion power plant development

Direct-drive inertial confinement fusion: A review

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

Contact Info

Product

Resources

About