skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

Title: Inertial fusion energy power reactor fuel recovery system

Abstract

A conceptual design is proposed to support the recovery of un-expended fuel, ash, and associated post-detonation products resident in plasma exhaust from a {approx}2 GWIFE direct drive power reactor. The design includes systems for the safe and efficient collection, processing, and purification of plasma exhaust fuel components. The system has been conceptually designed and sized such that tritium bred within blankets, lining the reactor target chamber, can also be collected, processed, and introduced into the fuel cycle. The system will nominally be sized to process {approx}2 kg of tritium per day and is designed to link directly to the target chamber vacuum pumping system. An effort to model the fuel recovery system (FRS) using the Aspen Plus engineering code has commenced. The system design supports processing effluent gases from the reactor directly from the exhaust of the vacuum pumping system or in batch mode, via a buffer vessel in the Receiving and Analysis System. Emphasis is on nuclear safety, reliability, and redundancy as to maximize availability. The primary goal of the fuel recovery system design is to economically recycle components of direct drive IFE fuel. The FRS design is presented as a facility sub-system in the context of supporting themore » larger goal of producing safe and economical IFE power. (authors)« less

Authors:
; ; ;  [1]; ;  [2];  [3]
  1. Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)
  2. Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)
  3. Savannah River National Laboratory, Aiken, SC 29808 (United States)
Publication Date:
OSTI Identifier:
22109378
Resource Type:
Journal Article
Resource Relation:
Journal Name: Fusion Science and Technology; Journal Volume: 54; Journal Issue: 2; Conference: 8. international conference on tritium science and technology, Rochester, NY (United States), 16-21 Sep 2007; Other Information: Country of input: France; 2 refs
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; BUFFERS; CONTROL; DESIGN; FUEL CYCLE; GAS FUELS; INERTIAL FUSION DRIVERS; NUCLEAR FUELS; PLASMA; POWER REACTORS; PURIFICATION; RELIABILITY; SAFETY; SYSTEMS ANALYSIS; TARGET CHAMBERS; TRITIUM; VACUUM PUMPS

Citation Formats

Gentile, C. A., Kozub, T., Langish, S. W., Ciebiera, L. P., Nobile, A., Wermer, J., and Sessions, K. Inertial fusion energy power reactor fuel recovery system. United States: N. p., 2008. Web.
Gentile, C. A., Kozub, T., Langish, S. W., Ciebiera, L. P., Nobile, A., Wermer, J., & Sessions, K. Inertial fusion energy power reactor fuel recovery system. United States.
Gentile, C. A., Kozub, T., Langish, S. W., Ciebiera, L. P., Nobile, A., Wermer, J., and Sessions, K. 2008. "Inertial fusion energy power reactor fuel recovery system". United States. doi:.
@article{osti_22109378,
title = {Inertial fusion energy power reactor fuel recovery system},
author = {Gentile, C. A. and Kozub, T. and Langish, S. W. and Ciebiera, L. P. and Nobile, A. and Wermer, J. and Sessions, K.},
abstractNote = {A conceptual design is proposed to support the recovery of un-expended fuel, ash, and associated post-detonation products resident in plasma exhaust from a {approx}2 GWIFE direct drive power reactor. The design includes systems for the safe and efficient collection, processing, and purification of plasma exhaust fuel components. The system has been conceptually designed and sized such that tritium bred within blankets, lining the reactor target chamber, can also be collected, processed, and introduced into the fuel cycle. The system will nominally be sized to process {approx}2 kg of tritium per day and is designed to link directly to the target chamber vacuum pumping system. An effort to model the fuel recovery system (FRS) using the Aspen Plus engineering code has commenced. The system design supports processing effluent gases from the reactor directly from the exhaust of the vacuum pumping system or in batch mode, via a buffer vessel in the Receiving and Analysis System. Emphasis is on nuclear safety, reliability, and redundancy as to maximize availability. The primary goal of the fuel recovery system design is to economically recycle components of direct drive IFE fuel. The FRS design is presented as a facility sub-system in the context of supporting the larger goal of producing safe and economical IFE power. (authors)},
doi = {},
journal = {Fusion Science and Technology},
number = 2,
volume = 54,
place = {United States},
year = 2008,
month = 7
}
  • In the High-Yield Lithium-Injection Fusion-Energy (HYLIFE) power plant design, lithium is replaced by molten salt. HYLIFE-II [Fusion Technol. {bold 25}, 5 (1994)] is based on nonflammable, renewable-liquid-wall fusion target chambers formed with Li{sub 2}BeF{sub 4} molten-salt jets, a heavy-ion driver, and single-sided illumination of indirect-drive targets. Building fusion chambers from existing materials with life-of-plant structural walls behind the liquid walls, while still meeting non-nuclear grade construction and low-level waste requirements, has profound implications for inertial fusion energy (IFE) development. Fluid-flow work and computational fluid dynamics predict chamber clearing adequate for 6 Hz pulse rates. Predicted electricity cost is reduced aboutmore » 30% to 4.4{cents}/kWh at 1 GWe and 3.2{cents}/kWh at 2 GWe. Development can be foreshortened and cost reduced by obviating expensive neutron sources to develop first-wall materials. The driver and chamber can be upgraded in stages, avoiding separate and sequential facilities. Important features of a practical IFE power plant are ignition and sufficient gain in targets; low-cost, efficient, rep-ratable driver; and low-cost targets.« less
  • Aneutronic fusion reactions can be considered as the cleanest way to exploit nuclear energy. However, these reactions present in general two main drawbacks: - very high temperatures are needed to reach relevant values of their cross sections; - moderate (and even low) energy yield per reaction. This value is still lower if measured in relation to the Z number of the reacting particles. It is already known that bremsstrahlung overruns the plasma reheating by fusion born charged-particles in most of the advanced fuels. This is for instance the case for proton-boron-11 fusion in a stoichiometric plasma and is also somore » in lithium isotopes fusion reactions. In this paper, the use of deuterium-tritium seeding is suggested to allow to reach higher burnup fractions of advanced fuels, starting at a lower ignition temperature. Of course, neutron production increases as DT contents does. Nevertheless, the ratio of neutron production to energy generation is much lower in DT-advanced fuel mixtures than in pure DT plasmas. One of the main findings of this work is that some natural resources (as D and Li-7) can be burned-up in a catalytic regime for tritium. In this case, neither external tritium breeding nor tritium storage are needed, because the tritium inventory after the fusion burst is the same as before it. The fusion reactor can thus operate on a pure recycling of a small tritium inventory.« less
  • Probably the single largest advantage of the inertial route to fusion energy (IFE) is the perception that its power plant embodiments could achieve acceptable capacity factors. This is a result of its relative simplicity, the decoupling of the driver and reactor chamber, and the potential to employ thick liquid walls. The author examines these issues in terms of the complexity, reliability, maintainability and, therefore, availability of both magnetic and inertial fusion power plants and compares these factors with corresponding scheduled and unscheduled outage data from present day fission experience. The author stresses that, given the simple nature of a fissionmore » core, the vast majority of unplanned outages in fission plants are due to failures outside the reactor vessel itself. Given one must be prepared for similar outages in the analogous plant external to a fusion power core, this puts severe demands on the reliability required of the fusion core itself. The author indicates that such requirements can probably be met for IFE plants. He recommends that this advantage be promoted by performing a quantitative reliability and availability study for a representative IFE power plant and suggests that databases are probably adequate for this task. 40 refs., 4 figs., 3 tabs.« less
  • In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain sizemore » should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.« less
  • The so-called 'threat spectra' of an inertial fusion energy (IFE) high gain target (neutron, x-ray, and ion energy fraction and particle spectra) are the usual starting point for IFE reactor conceptual design. The threat spectra are typically computed using the same radiation hydrodynamics and thermonuclear burn computer simulation codes used to compute implosion, ignition and burn. We analyze the validity of this model for simulating the expansion of the direct drive IFE target plasma and for computing threat spectra. Particular attention is paid to the collisionality of the expanding plasma.