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DOE OSTI2022

Fusion Blanket and Fuel Cycle Research at PNNL: FY22 Year-end Report

During the reporting period, research at PNNL focused on two tasks within the DOE Fusion Blanket and Fuel Cycle Program. Research on Task 1, Tritium Extraction from Pb-Li and He Using a Vacuum Permeator, focused on atomistic modeling to better understand tritium transport in Pd-coated V vacuum permeators. As a lower cost alternative to Pd permeators, thin coatings of Pd (or other noble metals) can be deposited over a substrate like V. However, the permeation performance of composite metal membranes degrades over time, due to the formation of intermetallics at the coating-substrate interface. Computational studies were performed to better understand tritium transport through these Pd-V intermetallics. The results of the FY22 Pd-V modeling study were recently submitted for publication in Computational Materials Science and presented at the Technology of Fusion Energy conference. Future work in this area will focus on interdiffusion barriers to prevent intermetallic formation that is deleterious to tritium transport. There are opportunities for collaboration with researchers at the Colorado School of Mines, who are manufacturing and testing candidate interdiffusion barriers. Research on Task 3, Solid Breeder Materials, included ion irradiation and post-irradiation characterization of lithium orthosilicate (Li 4 SiO 4 ) and lithium metasilicate (Li 2 SiO 3 ) to improve fundamental understanding of irradiation effects, in combination with atomistic modeling focused on the energetics of He clustering in these two ceramic phases. The results of the study suggested that the Li 4 SiO 4 phase, which is more desirable as a solid breeder due to its higher Li density, was amorphized during ion irradiation while the Li 2 SiO 3 phase appeared to be more resistant to irradiation damage. It is possible that Li loss contributed to the poor irradiation performance of the Li 4 SiO 4 , and some thoughts are provided regarding coatings that could be applied to solid breeders like this to prevent Li loss at elevated temperature while not hindering tritium diffusion. The results of the FY22 ion irradiation study were recently submitted for publication in Journal of Nuclear Materials and presented at the 22nd International Conference on Ion Beam Modification of Materials. Future work in this area will focus on Li-rich ceramics such as Li 5 AlO 4 and Li 8 ZrO 6 that have high Li density and should provide rapid tritium release based on previous work with less Li-rich ceramics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1992

Fusion energy from the Moon for the twenty-first century

It is shown in this paper that the D-He-3 fusion fuel cycle is not only credible from a physics standpoint, but that its breakeven and ignition characteristics could be developed on roughly the same time schedule as the DT cycle. It was also shown that the extremely low fraction of power in neutrons, the lack of significant radioactivity in the reactants, and the potential for very high conversion efficiencies, can result in definite advantages for the D-He-3 cycle with respect to DT fusion and fission reactors in the twenty-first century. More specifically, the D-He-3 cycle can accomplish the following: (1) eliminate the need for deep geologic waste burial facilities and the wastes can qualify for Class A, near-surface land burial; (2) allow 'inherently safe' reactors to be built that, under the worst conceivable accident, cannot cause a civilian fatality or result in a significant (greater than 100 mrem) exposure to a member of the public; (3) reduce the radiation damage levels to a point where no scheduled replacement of reactor structural components is required, i.e., full reactor lifetimes (approximately 30 FPY) can be credibly claimed; (4) increase the reliability and availability of fusion reactors compared to DT systems because of the greatly reduced radioactivity, the low neutron damage, and the elimination of T breeding; and (5) greatly reduce the capital costs of fusion power plants (compared to DT systems) by as much as 50 percent and present the potential for a significant reduction on the COE. The concepts presented in this paper tie together two of the most ambitious high-technology endeavors of the twentieth century: the development of controlled thermonuclear fusion for civilian power applications and the utilization of outer space for the benefit of mankind on Earth.

Kulcinski, G. L.↗

DOE OSTI2026

A Model of Hydrogen Solubility in Palladium-Silver Alloys

Fusion fuel cycle designs depend on palladium and palladium-silver alloys. The selective permeability of these is used to separate hydrogen isotopes from other elements. Predicting the solubility of hydrogen isotopes in palladium-silver alloys is important for the design of key unit operations for tritium processing. This paper presents a model of hydrogen isotope solubility in palladium and palladium-silver alloys based on Guggenheim’s quasichemical lattice theory. The model parameters were determined as functions of isotope molecular weight and the silver content of the alloy. The model predicts reasonable values for the solubility of protium, deuterium, and tritium over a wide range of temperatures, pressures, and alloy compositions.

08 HYDROGEN↗

DOE OSTI2020

Surface effects on deuterium permeation through vanadium membranes

Dense vanadium-based membranes offer high permeability and perfect selectivity to hydrogen isotopes, maintain favorable neutronic properties, and are compatible with liquid metals such as PbLi. These properties make vanadium membranes a promising fusion fuel cycle technology for processes such as tritium extraction from PbLi and exhaust processing. Surface contamination has a deleterious effect on the gas-phase hydrogen permeation through vanadium, and the reported permeabilities range from 10 -14 to 10 -7 mol m -1 s -1 Pa -0.5 . Thin dense films of palladium applied to clean vanadium surfaces enable a consistently high hydrogen permeability. In this study, uncoated vanadium resulted in deuterium permeabilities ranging from 2.8 × 10 -11 to 6.4 × 10 -9 mol m -1 s -1 Pa -0.5 at 300 °C–700 °C, respectively. Post-test analysis revealed a VO x surface layer and VC x subsurface layer formed on the feed side, while the as-received surface oxide dissolved leaving a submonolayer oxide on the permeate surface. Furthermore, the Pd-coated V resulted in a maximum deuterium permeability of 2.1 × 10 -7 m -1 s -1 Pa -0.5 at 375 °C upon activation of the Pd surface by oxidation and reduction. The deuterium permeation declined upon heating to 500 °C due to intermetallic diffusion between the Pd and V. The Mo 2 C-coated V resulted in deuterium permeabilities ranging from 2.7 × 10 -10 to 1.8 × 10-9 at 500 °C–700 °C, respectively, and a post-test analysis found the carbon in the Mo 2 C layer had dissolved into the V near the interface.

13 HYDRO ENERGY↗

DOE OSTI2025

Integrated operation scenarios: Chapter 6 of the special issue: on the path to tokamak burning plasma operation

Here we report the progress of the development and optimization of operational scenarios for ITER and beyond, focusing upon baseline, hybrid, and steady-state scenarios since 2007. This includes advancements made by the integrated operation scenarios (IOS) topical group of the international tokamak physical activity as well as contributions from the broader tokamak community. The key area of research involves developing IOSs that encompass tokamak physics, operation, and technology by utilizing integrated modeling and control strategies. This requires leveraging available actuators to simultaneously control plasma position and shape, MHD activities that could lead to disruptions, transport, plasma-wall interaction and power exhaust, fuel cycle, fusion burn, and tritium breeding. The control extends from the plasma initiation phase, through the current ramp-up, flattop, start and end of the fusion burn, and current ramp-down, to the plasma termination phase. A review of the currently developed scenarios and modeling is provided in terms of (i) optimizing plasma initiation in ITER, (ii) preparing for the low activation phase to fully commission all tokamak systems and establish and validate physics and scenario conditions in preparation for deuterim-tritium (DT) operation, (iii) developing and preparing baseline and hybrid scenarios to demonstrate the feasibility of achieving these regimes within device constraints, (iv) exploring steady-state scenarios to meet ITER’s steady-state goals, (v) evaluating and preparing actuators for ITER, (vi) developing integrated control solutions using shared actuators. The most notable achievements include; (i) the development of ITER demonstration discharges by matching various dimensionless parameters, (ii) the development of scenarios in an ITER-like tungsten environment and DT operation, and (iii) the development of scenarios in superconducting tokamaks, enabling long-pulse operations with similar coil constraints to ITER. Along with these significant achievements, outstanding issues and recommendations for further research and development are provided. Importantly, this study goes beyond simply updating the ITER Physics Basis; it carries profound implications for the broader field of burning plasma research, offering valuable insights and guidance for the next generation of fusion experiments and devices.

ITER↗

DOE OSTI2025

Effects of Target Protium Content on SteadyState Isotope Rebalancing and Protium Removal Distillation Column Operation

• SRNL Fusion Fuel Cycle Research • SRNL Fuel Cycle Tritium Inventory Optimization Approaches • RHINO/Aspen • CODFISH • Direct Internal Recycling • IFE Fuel Cycle Overview • Direct Internal Recycling Applied to IFE • CODFISH Isotope Rebalancing and Protium Removal Column Optimization • RHINO Fuel Cycle Analysis • Conclusions and Future Work

Somers, Alex [Savannah River National Laboratory (↗

DOE OSTI2025

Techno-Economic Analysis of Gas-Liquid Contactors for Tritium Extraction from Lead-Lithium

To enable a sustainable fuel cycle, any deuterium-tritium fusion reactor must breed its tritium fuel onsite. Lead-lithium (PbLi), a eutectic metal, is a leading liquid breeder material for tritium generation. One challenge with PbLi blanket technology is the extraction of tritium from the molten eutectic. Three technologies are the focus of worldwide research: the vacuum permeator, the vacuum sieve tray, and the gas-liquid contactor (GLC). The present work offers a methodology for designing, sizing, optimizing, and costing a trickle-bed GLC. Here, we analyzed tritium extraction from PbLi using MELODIE experimental data by applying traditional packed bed mass transfer efficiency models along with supplementary models, like film theory. Our analysis revealed that traditional packed bed mass transfer efficiency models do not fit the MELODIE loop experimental data. Moreover, uncertainty in PbLi solubility resulted in a 325-fold increase in required gas flow rates when comparing identical packing heights. The film theory liquid mass transfer coefficient, Delt-Olujic wettability model, and Reiter tritium solubility values fit the MELODIE data best and were used both in the design and to conduct the economic analysis. Techno-economic analysis of the GLC was performed to evaluate three design sizes, all achieving a minimum extraction efficiency of 90 [%] for a total tritium extraction of 31 [kg/yr].

Fusion Fuel Cycle↗

DOE OSTI2024

Deuterium Permeation Through a Self-Supported Palladium-Silver Membrane in Helium Gas Mixtures

Palladium membranes and membrane reactors can separate and purify tritium from impurities in the plasma exhaust processing system for the fusion energy fuel cycle. Membranes can also act as a continuous separation method to remove tritium from helium streams in the breeder blanket tritium extraction system, such as from the purge gas of solid breeders. To investigate the potential of these membranes for these applications, we performed a deuterium permeation campaign with a self-supported palladium-silver (Pd-25Ag wt%) tube of 0.15 m length, 76 µ m wall thickness, and 3.0 mm inner diameter. A gas mixture of 3.95% D 2 and a balance of He was delivered to the inside of the tube and D permeated radially outwards through the membrane into a vacuum chamber. Permeation experiments were conducted at 300 ° C, 350 ° C, 400° C, and 450°C, from 100 to 1000 sccm total flow rate, and with total pressures of 90, 150, 190, and 250 kPa. Further, deuterium permeation was consistently lower than predicted from diffusion-limited permeation models, thus we developed a transport model that included gas-phase mass transfer and surface reactions to model experimental results. The dissociation constant was optimized to fit the developed model to experimental data.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Low temperature hydrogen plasma permeation in palladium and its alloys for fuel recycling in fusion systems

Superpermeation of hydrogen isotopes through metal foils is a critical component for efficient fuel recycling in fusion power systems. In that context hydrogen permeation through foils of palladium and its alloys with silver and copper was studied at low temperature (60 - 200°C) under plasma exposure. These alloys differ significantly in both bulk and surface properties, and comparisons can provide mechanistic insights. Here, permeation was observed only during plasma operation, confirming the negligible contribution of molecular hydrogen to the observed flux. As-received foils required surface treatment to achieve top performance. For Pd and Pd75Ag25 an oxidation treatment increased permeation an order of magnitude, but proved unstable as this desirable surface was reduced under hydrogen plasma exposure. In contrast, an Ar plasma cleaning step provided both high and stable flux. As-received FCC phase Pd60Cu40 foils required annealing to transform it into the high permeability BCC phase that delivered top performance. All foils displayed similar temperature dependence with flux declining with temperature, suggesting that the primary rate-limiting step is absorption of superthermal hydrogen. Among these foils the hydrogen flux through PdCu was 3 - 5X greater than that of Pd or PdAg, which were similar. The superiority of BCC PdCu is attributed to its superior hydrogen desorption kinetics. Using PdCu 100% permeation of supplied hydrogen was achieved, and the flux saturated with increasing plasma power at values >10 –2 mol H•m –2 •s– 1 , and under these conditions permeation rates are equivalent with or without the membrane present. The fluxes achieved are the highest reported to date at these conditions, and the results highlight the important roles of both surface and bulk properties.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Low temperature hydrogen superpermeation in vanadium composite metal foil pumps

Palladium-based foil membranes are an effective option for hydrogen isotope recovery from the plasma exhaust of future fusion plants, but cost and availability are concerns. Vanadium (V) is a relatively low cost, neutron tolerant material with high hydrogen permeability. It has been well-studied as a superpermeable membrane at high temperature (>500 °C), but V displays negligible superpermeation at low temperature (75–200 °C) due to catalytic limitations. Composite membranes were fabricated by depositing thin layers (~100 nm) of either Pd or BCC PdCu on sputter-cleaned vanadium foils (100 μm). Symmetric membranes elevated superpermeation to levels approaching bulk Pd or PdCu foils, with ~5X higher flux in the latter reflecting the superior properties of PdCu. Asymmetric membranes revealed that the Pd-based catalyst layer was critical for both efficient absorption of superthermal hydrogen upstream as well as catalyzing re-combinative desorption downstream. At T ≥ 150 °C composite membrane superpermeation was equivalent to the Pd-based foils, but the flux was attenuated by a factor of 2-3X as the temperature was reduced. This deviation from pure foil performance coincided with the formation of vanadium hydride (β-V 2 H), which also impacted the transient response. Nevertheless, no embrittlement was observed under the conditions examined and elevating the temperature >150 °C removed the hydride and restored full performance. The achievement of palladium-level performance with a >99% reduction in Pd inventory makes these V composite metal foils pumps an attractive option for low temperature hydrogen isotope recovery in future fusion plants.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Uranium Bed Design Parameters for Tritium Plants Supporting Fusion Reactors

Uranium hydride is commonly used to store hydrogen or its isotopes in a solid state. The Self-Assaying Tritium Accountancy and Containment Unit for ITER (STACI) is a 5.2-kg bed of depleted uranium (dU) capable of holding up to 33 mol of hydrogen or its isotopes. Here we summarize the data analysis of past experimental campaigns with STACI, with the aim of describing the kinetics and thermodynamics of the hydriding process. Computed tomography imaging was performed on STACI both before and after its experimental campaign, and a high degree of swelling was observed in the dU. Literature on studies in regard to the swelling of large (multikilogram) quantities of uranium hydride for storage applications was not identified during this study. Data from the experimental campaign, as well as data on the formation reaction, are presented. The authors hope to create an analytical model of STACI based on these data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Continued Testing and Modeling of the AirSquared All-Metal Tritium Scroll Pump

The implementation of fusion energy requires the processing of the deuterium-tritium mixture used to fuel the reaction. Many gas processing operations require either a gas-motive force or a pressure gradient across a system; both are achieved using vacuum pumps. However, because of the material challenges of working with tritium, specialized pumps are required that use only metal on wetted surfaces. This prevents tritium retention and degradation of materials such as oils and polymers and reduces the generation of tritiated waste. Typical gas handling systems for tritium operations involve metal bellows pumps, all-metal scroll pumps, and metal turbomolecular pumps, depending on the pressure requirements. With the industry standard Normetex Type 15 m 3 /h (typically referred to as the Normetex 15) pump no longer available, the fusion market requires viable alternatives. In this study, we test an American-made pump, the AirSquared V16H034A-C01, and compare its performance to the Normetex. A performance model of the AirSquared pump is also produced, which can be applied to general process models to generate an accurate description of pressures and flows through various unit operations.

08 HYDROGEN↗

DOE OSTI2022

Developments and Challenges in the Design of the ITER DRGA

We report the ITER Diagnostic Residual Gas Analyzer (DRGA) will measure the distribution of gas species, i.e., deuterium (D), tritium (T), and impurities, in the divertor exhaust stream and in the plasma periphery, with time resolution relevant to fusion plasma–wall particle dynamics. The uniqueness of the DRGA, over previous implementations of plasma dynamics residual gas analysis, is an integrated approach, combining mass and low-temperature plasma-activated optical spectroscopy, in a differentially pumped analysis station. A further unique feature of the ITER divertor-specific DRGA is an ~8-m separation of the analysis station from the sampled pumping duct, while still maintaining a ~1-s response time for hydrogen isotopic concentrations. ITER DRGA final design activities are strongly benefiting from testing of prototypical DRGA components and methods on present fusion devices, most currently on JET and W7-X. DRGA systems are implemented on both these devices and include sensors (and pumping methods) that are directly relevant to the ITER DRGA design. The recent JET-DTE2 campaign has provided the first experience on operating the combined ITER DRGA sensors with D-T plasmas. While enhancing system design for ITER, this experience on operating devices has also revealed additional engineering challenges, which further guide the continuing final design project. Meanwhile, the recent determination that the ITER DRGA, with slight optimization, will resolve the helium isotopes well enough to support an ITER pre-DT, He-3-based heating scheme, has greatly increased ITER Research Program interest in the DRGA and its implementation well ahead of the DT phase.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

Modeling and analysis of the tritium fuel cycle for ARC- and STEP-class D-T fusion power plants

Abstract The limited tritium resources available for the first fusion power plants (FPPs) make fuel self-sufficiency and tritium inventory minimization leading issues in FPP design. This work builds on the model proposed by Abdou et al (2020 Nucl. Fusion 61 013001), which analyzed the fuel cycle (FC) of a DEMOnstration nuclear FPP-class FPP with a time-dependent system-level model. Here, we use a modified version of their model to analyze the FC of an Affordable, Robust, Compact (ARC)-class tokamak and two versions of a Spherical Tokamak for Energy Production (STEP)-class tokamak. The ARC-class tokamak breeds tritium in a 2LiF + BeF 2 liquid immersion blanket, while the STEP-class tokamak breeds tritium utilizing either a liquid-lithium blanket design or an encapsulated breeding blanket. A time-dependent system-level model is developed in Matlab Simulink ® to simulate the evolution of tritium flows and tritium inventories in the FC. The main goals of this work are to assess tritium self-sufficiency of the ARC- and STEP-class designs and to determine quantitative design requirements that can be used to analyze the adequacy of a proposed FC system. These design requirements are aimed at achieving a low tritium inventory doubling time ( t d ) and a low start-up inventory ( I s t a r t u p ) while keeping the required tritium breeding ratio (TBR r ) as low as possible. We also consider how improvements in FC technology and POs affect TBR r and I s t a r t u p . The model results show that TBR r for ARC- and STEP-class FPPs should be achievable if the tritium burn efficiency (TBE) reaches 0.5%–1% (TBR r < 1.2). This assumes significant, but attainable, improvements over current abilities. However, the model results indicate that an FPP must achieve ambitious performance targets, including FPP availability > 70%, tritium processing time < 4 h, and the implementation of direct internal recycling (DIR). If future research yields major improvements to achievable TBE, it may be possible to achieve tritium self-sufficiency while operating at lower availability and without implementing DIR.

Physics↗

DOE OSTI2021

Natural hot-ion modes in a rotating plasma

We report that in steady state, the fuel cycle of a fusion plasma requires inward particle fluxes of fuel ions. These particle flows are also accompanied by heating. In the case of classical transport in a rotating cylindrical plasma, this heating can proceed through several distinct channels depending on the physical mechanisms involved. Some channels directly heat the fuel ions themselves, whereas others heat electrons. Which channel dominates depends, in general, on the details of the temperature, density and rotation profiles of the plasma constituents. However, remarkably, under relatively few assumptions concerning these profiles, if the α particles, the by-products of the fusion reaction, can be removed directly by other means, then a hot-ion mode tends to emerge naturally.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Comparison of Fuel Cycles for Lead-Lithium and Pure Lithium Liquid Metal Walls in a Magnetized Target Fusion Power Plant

General Fusion (GF) is developing an adaptable, commercial fusion power plant based on magnetized target fusion (MTF). The GF approach involves forming a spherical torus of deuterium-tritium plasma in a large (~4 m diameter) cavity formed in liquid metal, and then collapsing that cavity with an array of pneumatic piston drivers. The liquid metal is constantly flowing through the fusion chamber and out to processing systems where tritium and heat will be extracted using tritium extraction technologies and heat exchangers, respectively. Here, this study focuses on two candidate designs for the liquid metal blanket and first wall material for the General Fusion Magnetized Target Fusion (GF MTF) power plant and assesses their impact on the tritium fuel cycle. The first candidate is the lead lithium eutectic (LLE) and the second candidate is pure lithium (Li). It was found that the main differences between LLE and Li designs are the extraction technologies required to remove tritium from the blanket and the amount of tritium and its distribution within the facility. More than 80% of the in-process tritium inventory for the LLE design is contained in the isotope separation system, while for the Li design, over 60% of the in-process tritium inventory is contained within the blanket material. This is due to significant tritium retention by Li. For the Li blanket, the burden of tritium processing rests on the blanket extraction technology rather than the traditional exhaust processing route. Thus, the blanket extraction technology is a main driver of tritium inventory in the Li system and determines the subsequent interface with the tritium processing plant.

General Fusion↗

DOE OSTI2025

EM-Enhanced HyPOR Loop for Fast Fusion Cycles

Vacuum pumps are the heart of a fusion energy facility – fusion power cannot be generated without them. Nevertheless, the vacuum technology needed to operate a viable fuel cycle for a compact fusion power plant does not exist. Commercial vacuum technology offers the best solution to this challenge, but a pump oil recycling and detritiation system is necessary. Conventional oil detritiation processes have only been developed to deal with legacy waste and are too slow and destructive to the oil. Further, post hoc detritiation strategies are intrinsically inefficient. Our approach is to rethink the challenge holistically by designing the pumping fluid and tritium extraction system in an integrated manner to achieve an innovative solution. By selecting an oil for the specific task of pumping tritium and then designing our catalytic system to selectively target the tritium-bearing functional groups, an effective detritiation system (hydrocarbon pump oil recycling, HyPOR, loop) for fusion power plants can be obtained. The project demonstrated a HyPOR loop process that can selectively remove heavier hydrogen isotopes from pump oil, reaching the target of 99.5 % removal, with an uptake of less than 0.01% of tritium throughput, while also purifying the oil of radiation-induced damage. The recycled oil retained its pumping characteristics over 7 recycles and gamma irradiation over 30 MGy. By meeting these targets, the project enable a reduction in pump operational costs from $\$$14.5M/year to $\$$103k/year (>140× reduction), reducing pump electric power consumption from 2.8 MW to 0.25 MW (>10× reduction), and reducing in-process tritium inventory from 2.03 kg to 0.48 kg (>4× reduction).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

NASA NTRS1988

Energy requirements for HE-3 mining operations on the Moon

At the present rate of world energy consumption (10 TW-y/y) and allowing for an equilibrium consumption of 20 to 30 TW-y/y in mid 21st century, we will exhaust economically recoverable fossil fuels in the next 50 to 60 years. We will then have to rely on nuclear (fission and fusion) and renewable energy to feed, warm, and protect the world's population. Fusion energy is expected to play an important role in the 21st century and there a 2 billion dollar per year research program to commercialize that energy resource. A serious problem with this is its reliance on the D-T fuel cycle which releases 80 percent of its energy in the form of neutrons. These neutrons cause significant radiation damage and induce large amounts of radioactivity. There is another fusion fuel cycle involving the isotopes of Deuterium and Helium-3 which, if configured properly, releases 1 percent or less of its energy in neutrons. Obviously, such a fuel would be preferred, but there is no large source of He-3 known to satisfy world energy needs. Fortunately, a very large source of He-3 was found on the Moon, implanted over the past 4 billion years by the solar wind. Recent analysis of Apollo and Luna data reveals that over a million tons of He-3 sit on the Moon's surface. The potential energy in this He-3 fuel is 10 times that contained in all the coal, oil, and natural gas on the Earth. The purpose of this paper is to examine the energy required to extract the He-3 from the lunar regolith.

Kulcinski, Gerald L.↗