Reaction balance and efficiency analysis of a D-D fusion/fission hybrid with satellite D-3He reactors

Selected reactor physics and isotope balance characteristics of a fusion hybrid supported D-³He satellite nuclear energy system are formulated and investigated. The system consists of two types of reactors: a parent D-fueled fusion device and a number of smaller reactors optimized for D-³He fusion. The parent hybrid station breeds the helium-3 for the satellites and also breeds fissile fuel for an existing fission reactor economy. Various hybrid operational regimes are examined in order to determine favorable reactorQ values and effective fusion and fission efficiencies. A number of analytical correlations between power output, plasma energetics, blanket neutronics, breeding capacity, and energy conversion cycles are established and evaluated. Numerical examples of performance parameters such as fission-to-fusion power, overall conversion efficiency, and the ratio of satellite to parent fusion power are presented. The range of reactor efficiencies is elucidated as affected by the internal plasma power balances. As an upper bound based on optimistic injection and direct conversion efficiencies, we find the D-³He satellite system power output attaining at best 1/3 of the parent fusion power.     

Optimization of nonthermal fusion power consistent with channeling of charged fusion product energy

If the energy of charged fusion products can be diverted directly to fuel ions, non-Maxwellian fuel ion distributions and temperature differences between species will result. To determine the importance of these nonthermal effects, the fusion power density is optimized at constant- for nonthermal distributions that are self-consistently maintained by channeling of energy from charged fusion products. For D-T and D-³He reactors, with 75% of charged fusion product power diverted to fuel ions, temperature differences between electrons and ions increase the reactivity by 40–70%, while non-Maxwellian fuel ion distributions and temperature differences between ionic species increase the reactivity by an additional 3–15%.     

Exploratory studies of tokamaks as fusion test reactors

Studies have been performed to explore various plasma burn scenarios for a tokamak test reactor which could follow the next generation of large tokamak experiments. Tradeoffs between an ignited burning plasma and a sub-ignited driven plasma are examined in terms of device size and performance as a fusion engineering test facility. It is found that plasma performance levels, measured by ignition margin, amplification factorQ, and fusion power output, increase with device size, more optimistic transport scaling laws, lower magnetic field ripple, and higher. The performance of a generally low stress (B ₀=4 T) reference device, with major radiusR=4.5 m and minor radiusa=1.3 m in a D-shaped (=1.6) plasma has been evaluated over a wide range of operating parameters. In particular, a moderate fusion power output of 300 MW is obtained, the driven plasma havingQ 10, an edge ripple of 1%, and a density ranging between 1.0 and 1.5×10¹⁴ cm^–3. The same device operated at a higher general level of stress (B ₀=5.3 T) is predicted to achieve ignition, but is not required for the mission of an engineering test facility and would entail greater technical risk.     

Advantages of high-field tokamaks for fusion reactor development

High-field designs could reduce the cost and complexity of tokamak reactors. Moreover, the certainty of achieving required plasma performance could be increased. Strong Ohmic heating could eliminate or significantly decrease auxiliary heating power requirements and high values of n_E could be obtained in modest-size plasmas. Other potential advantages are reactor operation at modest values of , capability of higher power density and wall loading, and possibility of operation with advanced fuel mixtures. Present experimental results and basic scaling relations imply that the parameterB ²a, where B is the magnetic field and a is the minor radius, may be of special importance. A superhigh-field compact ignition experiment with very high values ofB ²a (e.g.,B ²a=150 T² m) has the potential of Ohmically heating to ignition. This short-pulse device would use inertially cooled copper plate magnets. Compact engineering test reactor and/or experimental hybrid reactor designs would use steady-state, water-cooled copper magnets and provide long-pulse operation. Design concepts are also described for demonstration/commercial reactors. These devices could use high-field superconducting magnets with 7–10 T at the plasma axis.     

Lessons learned from recent D-3He tokamak power reactor studies

Conclusions In conclusion, the advantages of the D-³He fuel cycle are compelling but the challenges are great. The economics of a fusion power plant with a permanent first wall, especially in terms of availability and reliability, should translate into an attractive future option for society. The safety and environmental features of this type of power could make this energy source irresistible for a world choked by pollution and racked with wars over the remaining scraps of fossil fuel energy. These advantages will not come free; the physics in terms of highernT, larger plasma currents, ash removal requirements, and the need to use fuel from settlements on the Moon (which should be in place long before we need³He for power reactors) are all problems that need to be solve in the next 20 years. The benefits to mankind surely will overweigh these latter problems and the nation, or nations who develop this energy source will have an important strategic advantage in the twenty-first century.     

Recent developments in environmental aspects of D-He fuelled fusion devices

The stress on fusion safety has stimulated worldwide research in the late 1980s for fuel cycles other than D-T. With advanced cycles, such as D-D, D-³He, p-¹¹B, and ³He-³He, it is not necessary to breed and fuel large amounts of tritium. The D-³He fuel cycle in particular is not completely aneutronic due to the side D-D reactions. Neutron wall loadings, however, can be kept low (by orders of magnitude) compared to D-T fuelled plants with the same output power, eliminating the need for replacing the first wall and shielding components during the entire plant lifetime. Other attractive safety characteristics include low activity and decay heat levels, low-level waste, and low releasable radioactive inventory from credible accidents.There is a growing international effort to alleviate the environmental impact of fusion and to support the most recent trend in radwaste management that suggests replacing the geological disposal option with more environmentally attractive scenarios, such as recycling and clearance. We took the initiative to apply these approaches to existing D-³He conceptual designs: the ARIES-III power plant and the Candor experiment. Furthermore, a comparison between the radiological aspect of the D-³He and D-T fuel cycles was assessed and showed notable differences. This report documents the comparative assessment and supports the safety argument in favour of the D-³He fuel cycle.     

Experimental results of the erosion of molybdenum due to blistering by He+ bombardment

We report some preliminary measurement of the erosion rate of plasma dumps when bombarded with 100 keV He⁺ ions at high power density ( 1 MW/m²). The expected erosion rates, based on measurements of He blistering that were made at lower power density ( 0.3 MW/m²), indicate a potentially serious problem for fusion reactors. Our tests use a reactorlike power density and produce He blisters at a rate that is slower than predicted by 2 to 4 orders of magnitude, depending on the temperature of the molybdenum target.     

Nuclear fuels for low-beta fusion reactors: Lithium resources revisited

In searching to attain optimum conditions for the controlled release of nuclear energy by fusion processes, the stationary confinement of low-pressure ring-shaped plasmas by strong magnetic fields is now regarded as the most promising approach. We consider a number of fuel combinations that could be operated in such low-beta reactor systems and look upon the relevant fuel reserves. The classical D-T-Li cycle will be used as a standard and is extensively discussed therefore. It could supply most of mankind's future long-term power needs—but only on condition that the required lithium fuel can be extracted from seawater at reasonable expenses. The estimated landbound lithium reserves are too small to that end, they will last for about 500 years at most, depending on forecasts of future energy consumption and on assumptions about exploitable resources. Recovery of lithium from seawater would extend the possible range by a factor of 300 or so, provided that extraction technologies which are at present available in the laboratory, could be extended to a very large and industrial scale. Deuterium is abundant on earth but D-D fusion is difficult, if not impossible, to be achieved in the low-beta systems presently investigated for D-T fusion. The same arguments apply to so-called advanced concepts, such as the D-³He and the D-⁶Li cycles.     

Neutronic analysis of an inertial fusion energy breeder with SiCf/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC_f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ or ⁴He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li₁₇Pb₈₃ coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). ⁴He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B₄C shielding.     

Design study of a 120-keV,3He neutral beam injector

We describe a design for a 120-keV, 2.3-MW,³He neutral beam injector for use on a D-³He fusion reactor. The constraint that limits operating life when injecting He is its high sputtering rate. The sputtering is partly controlled by using an extra grid to prevent ion flow from the neutralizer duct to the electron suppressor grid, but a tradeoff between beam current and operating life is still required. Hollow grid wires functioning as mercury heat pipes cool the grid and enable steady state operation. Voltage holding and radiation effects on the acceleration grid structure are discussed. We also briefly describe the vacuum system and analyze use of a direct energy converter to recapture energy from unneutralized ions exiting the neutralizer. Of crucial importance to the technical feasibility of the³He-burning reactor are the injector efficiency and cost; these are 53% and $5.5 million, respectively, when power supplies are included.The beam is composed of 91 separate, parallel currents that flow in the gaps between the elements or wires of a grid. Each such flow is referred to as a beamlet. The current densities in Figs. 5, 8, and 9 are values within a beamlet, as measured at the beam-forming grid. They are not values averaged over the entire beam cross-section.     

Could Advanced Fusion Fuels Be Used with Today's Technology?

Could today's technology suffice for engineering advanced-fuel, magnetic-fusion power plants, thus making fusion development primarily a physics problem? Such a path would almost certainly cost far less than the present D-T development program, which is driven by daunting engineering challenges as well as physics questions. Advanced fusion fuels, in contrast to D-T fuel, produce a smaller fraction of the fusion power as neutrons but have lower fusion reactivity, leading to a trade-off between engineering and physics. This paper examines the critical fusion engineering issues and related technologies with an eye to their application in tokamak and alternate-concept D-³He power plants. These issues include plasma power balance, magnets, surface heat flux, input power, fuel source, radiation damage, radioactive waste disposal, and nuclear proliferation.     

High power pulsed lasers

Pulsed high power lasers can deliver sufficient energy on inertial fusion time scales (0.1–10 ns) to heat and compress DT fuel to fusion reaction conditions. Several laser systems have been examined for application to the fusion problem. Examples are Ndglass, CO₂, KrF, and I₂, etc. A great deal of developmental effort has been applied to the Ndglass laser and the CO₂ gas laser systems. These systems now deliver >10⁴ kJ and >20×10¹² W to inertial fusion targets. The Nova Ndglass laser is being constructed to provide >200 kJ and >200×10¹² W of 1 m radiation for fusion experimentation in the mid-1980s. For inertial fusion target gain, >100 times the laser input, it is expected that the laser must deliver 3–5 MJ of energy on the 10–20 ns time scale. This paper reviews the developments in laser technology and outlines approaches to construction of a 3–5 MJ driver.     

The reversed-field-pinch (RFP) fusion neutron source: A conceptual design

The conceptual design of an ohmically heated, reversed-field pinch (RFP) operating at 5-MW/m² steady-state DT fusion neutron wall loading and 124-MW total fusion power is presented. These results are useful in projecting the development of a cost effective, low-input-power (206 MW) source of DT neutrons for large-volume (10 m³), high-fluence (3.4 MW yr/m²) fusion nuclear materials and technology testing.Work supported by U.S. DOE.     

Compact ignition test reactor (CITR)

Design considerations have been developed for a compact ignition test reactor (CITR). The objectives of this tokamak device are to achieve ignition, to study the characteristics of plasmas that are self-heated by alpha particles, and to investigate burn control. To achieve a compact design, the toroidal field magnet consists of copper-stainless steel plates to accommodate relatively high stresses; it is inertially cooled by liquid nitrogen. No neutron shielding is provided between the plasma and the toroidal field magnet. The flat-top of the toroidal field magnet is 10 s. Strong auxiliary heating is employed. In one design option, adiabatic compression in major radius is employed to reduce the neutral beam energy required for adequate penetration; thiscompression boosted design option has a horizontally elongated vacuum chamber; illustrative parameters are a compressed plasma witha=0.50 m, R=1.35 m,B _T =9.1 T, and a neutral beam power of 15 MW of 160 keVD ⁰ beams. A design option has also been developed for alarge bore device, which utilizes a circular vacuum chamber. Thelarge bore design provides increased margin and flexibility; both direct heating with RF or neutral beam injection and compression boosted startup are possible. The large bore design also facilitates the investigation of high-Q driven operation. Illustrative plasma parameters for full use of the large bore area=0.85 m,R=1.90 m, andB _T =7.5 T.     

Inertial Confinement Fusion with a Heavy-Ion Driver-Accelerator and a Cylindrical Target

An economically efficient power plant burning deuterium-tritium fuel can be built using a powerful heavy-ion accelerator of a new type. A multilinear cryogenic cylindrical target, 1 cm long, 0.44 cm in radius, and containing 7.8 mg equimolar DT fuel, is studied as an example. The driver-accelerator gives two different target irradiation regimes. In both regimes, the beams consist of platinum ions, accelerated up to 100 GeV, with different isotopic composition, charged in the first regime only positively and in the second regime positively and negatively. In the first regime, the beam energy is 4.8 MJ and the beam heats in 60 nsec only the target shell. High heating symmetry is achieved by rapidly rotating the beam around the target axis with frequency 1 GHz. The fuel is compressed into a dense filament, where the condition for propagation of a fusion burn wave is satisfied – R 0.4 g/cm². In the second regime, a beam with 0.4 MJ ions heats in 0.2 nsec compressed fuel with power density 2.5·10⁷ TW/cm² up to ignition temperature. The computed energy amplification factor in the target is 200.     

The need and prospects for improved fusion reactors

Conceptual fusion reactor studies over the past 10–15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100–200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed.Nomenclature a Plasma minor radius at outboard equatorial plane (m) - A Plasma aspect ratioR _T /a - AC Annual charges ($/yr) - b Plasma minor radius in vertical direction (m) - B Magentic field at plasma or blanket (T) - B _c Magnetic field at the coil (T) - B Toroidal magnetic field (T) - B Poloidal magnetic field (T) - BOP Balance of plant - C Coil - COE Cost of electricity (mills/kWeh) - CRFPR Compact RFP reactor - CT Compact torus (FRC or spheromak) - c _FPC Unit cost of fusion power core ($/kg) - DC Direct cost ($) - DZP Dense Z-pinch - E Escalation rate (1/yr) - EDC Escalation during construction ($) - ET Elongated tokamak - F Annual fuel charges ($/yr) - FC Component of UDC not strongly dependent or FPC size ($/kWe) - FW First wall - FPC Fusion power core - f _Aux Fraction of gross electric power recirculated to BOP - f ₁ (IC+IDC+EDC)/DC - f ₂ (O&M + SCR + F)/AC - IC Indirect cost ($) - IDC Interest during construction ($) - I _w Neutron first-wall loading (MW/m²) - i Toroidal plasma current (MA) - j Plasma current density, I/a² - k _B Boltzmann constant, 1.602(10)^–16 (J/keV) - LWR Light-water (fission) reactor - MPD Mass power density 1000P_E/M_FPC (kWe/tonne) - M _N Blanket energy multiplication of 14.1-MeV neutron energy - M _FPC Mass of fusion power core (tonne) - n Plasma density (m^–3) or toroidal MHD mode number - O&M Annual operating and maintenance cost ($/yr) - p _f Plant availability factor - PFD Poloidal field dominated (CTs, RFP, DZP) - P Construction time (yr) - P_TH Thermal power (MWt) - P _E Net electric power (1-)P _ET (MWe) - P_ET Total gross electric power (MWe) - p_f Fusion power (MW) - q Tokamak safety factor (B /B _gq )(a/R _T ) - q _e EngineeringQ value, 1/e - R _T Major toroidal radius (m) - RFP Reversed-field pinch - RPE Reactor plant equipment (Account 22) - S Shield - SCR Annual spare component cost ($/yr) - SSR Second stability region for the tokamak - S/T/H Stellarator/torsatron/heliotron - ST Spherical tokamak or spherical torus - T Plasma temperature (keV) - TDC Total direct cost ($) - TOC Total overnight cost ($) - UDC Unit direct cost,TDC/10 ³ P _E ($/kWe) - V _p Plasma volume (m³) - W _p Plasma energy (GJ) - W _B Magnetic field energy (GJ) - Magnetic utilization efficiency, 2nk_BT/(B ²/2₀) - ₀ Permeability of free space, 4(10)^–7 H/m - XE Plasma confinement efficiency, a²/4_E - _e Plasma energy confinement time - _p Overall plant efficiency, _TH(1-) - _TH Thermal conversion efficiency - _FPC AverageFPC mass density (tonne/m³) - Plasma vertical elongation factor,b/a - Thickness of allFPC engineering structure surround plasma (m) - Total recirculating power fraction, (P _ET-P _E)/P _ET, or inverse aspect ratioa/R _T This work was performed under the auspices of USDOE, Office of Fusion Energy.     

Transient thermo-mechanical behavior of a direct-drive target during injection in an inertial fusion energy chamber

During injection in an inertial fusion energy (IFE) chamber, a direct-drive target is subject to heat loads from chamber wall radiation and energy exchange from the chamber gas constituents. These heat loads can lead to the deuterium-tritium (DT) reaching its triple point temperature and even undergoing phase change, leading to unacceptable non-uniformity based on target physics requirements for compression and ignition of the DT fuel pellets using multiple laser beams. A two-dimensional bubble nucleation mode was added to the previously presented thermo-mechanical model to help better define the design margin for direct-drive IFE targets. The new model was validated by comparison with analytical results for controlled cases. It was then used to simulate heating experiments on DT targets conducted at the Los Alamos National Laboratory (LANL), where the ³He present in the DT due to tritium decay was found to affect the nucleation process.The previous requirement for target survival was for the temperature of the DT to remain below triple point of DT (19.79 K). If the existence of a melt layer does not violate the symmetry requirements on the target for successful implosion, the constraint could be relaxed by assuming a limit based on the avoidance of bubble nucleation. This study shows that the thresholds for melting and bubble nucleation are significantly different, allowing for extra margin in target survival under this assumption.     

Study of Fuel Ratios on the Fusion Reactivity in an Inertial Electrostatic Confinement Device Using a Residual Gas Analyzer

Gridded Inertial Electrostatic confinement (IEC) devices are of interest due to their flexibility in burning advanced fuels, their tuning ability of the applied voltage to the reaction cross-section. Although this device is not suitable for power production in its present form, it does have several near term applications. The number of applications of this device increases with increasing fusion reactivity. These devices are simple to operate but are inherently complicated to understand and an effort to incrementally understand the device to improve its operational efficiency is underway at University of Wisconsin, Madison. Of all the parameters under study we are focusing on the effects of flow rate and flow ratio on the fusion reactivity in the present paper. Experiments were conducted to understand the influence of fuel flow ratio on the fusion reactions. The residual gas analyzer (RGA) was used to study the impurity concentration as the flow ratio was changed. It was observed that the higher flow rate resulted in reduced impurity levels and hence an increase in fusion rate. Several different species of gases were detected, some of these molecules formed inside the RGA analyzer. The flow ratio scan revealed that the optimum mixture of D₂ with ³He to be D₂:³He::1:2 for maximum D–³He fusion rate.     

The high-density Z-pinch as a pulsed fusion neutron source for fusion nuclear technology and materials testing

The dense Z-pinch (DZP) is one of the earliest and simplest plasma heating and confinement schemes. Recent experimental advances based on plasma initiation from hair-like (10s m in radius) solid hydrogen filaments have so far not encountered the usually devastating MHD instabilities that plagued early DZP experimenters. These encouraging results along with the debut of a number of proof-of principle, high-current (1–2 MA in 10–100 ns) experiments have prompted consideration of the DZP as a pulsed source of DT fusion neutrons of sufficient strength (S _N 10¹⁹ n/s) to provide uncollided neutron fluxes in excess ofI _w = 5–10 MW/m² over test volumes of 10–30 liters or greater. While this neutron source would be pulsed (100s ns pulse widths, 10–100 Hz pulse rate), giving flux time compressions in the range 10⁵–10⁶, its simplicity, near-term feasibility, low cost, high-Q operation, and relevance to fusion systems thatmay provide a pulsed commercial end-product, e.g., inertial confinement or the DZP itself, together create the impetus for preliminary consideration as a neutron source for fusion nuclear technology and materials testings. The results of a preliminary parametric systems study (focusing primarily on physics issues), conceptual design, and cost vs. performance analyses are presented. The DZP promises an inexpensive and efficient means to provide pulsed DT neutrons at an average rate in excess of 10¹⁹ n/s, with neutron currents I_w10 MW/m² over volumes V_exp 30 liter using single-pulse technologies that differ little from those being used in present-day experiments.Work supported by U.S. DOE.