Numerical Study on Effects of Fuel Mixture Fraction and Li-6 Enrichment on Neutronic Parameters of a Fusion–Fission Hybrid Reactor

This study presents the effects of mixture fractions of nuclear fuels (mixture of fissile–fertile fuels and mixture of two different fertile fuels) and ⁶Li enrichment on the neutronic parameters (the tritium breeding ratio, TBR, the fission rate, FR, the energy multiplication ratio, M, the fissile breeding rate, FBR, the neutron leakage out of blanket, L, and the peak-to-average fission power density ratio, Γ) of a deuterium–tritium (D–T) fusion neutron-driven hybrid blanket. Three different fertile fuels (²³²Th, ²³⁸U and ²⁴⁴Cm), and one fissile fuel (²³⁵U) were selected as the nuclear fuel. Two different coolants (pressurized helium and natural lithium) were used for the nuclear heat transfer out of the fuel zone (FZ). The Boltzmann transport equation was solved numerically for obtaining the neutronic parameters with the help of the neutron transport code XSDRNPM/SCALE4.4a. In addition, these calculations were performed by also using the MCNP4B code. The sub-limits of the mixture fractions and ⁶Li enrichment were determined for the tritium self-sufficiency. The considered hybrid reactor can be operated in a self-sufficiency mode in the cases with the fuel mixtures mixed with a fraction of equal to or greater than these sub-limits. Furthermore, the numerical results show that the fissile fuel breeding and fission potentials of the blankets with the helium coolant are higher than with the lithium coolant.     

The Spallator and APEX Nuclear Fuel Cycle a New Option for Nuclear Power

A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high energy (1 to 2 GeV) protons on a heavy metal target. The neutrons are absorbed in a surrounding natural uranium or thorium blanket in which fissile Pu-239 or U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high beam current continuous wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of the short-lived fission products external to the fuel cycle eliminates the need for long-term geological age storage of fission product waste.     

Suitability of Small Scale Linear Systems for a Fission–Fusion Reactor, Breeder, and Waste Transmutation

To date the magnetic fusion effort has been almost entirely devoted to only one application, that being a multi gigawatt central station electric plant. Given the already well established fission based industry, the likelihood that fusion will have any impact on curbing the current carbon-based energy demand in the 21st century is slim. This paper advocates that the first and primary use of fusion neutrons should be as the driver for a sub-critical fission nuclear energy system—a fission–fusion hybrid reactor. This system can also be utilized to transmute long-lived radioactive wastes, and breed fissile nuclear fuel for several additional fission reactors. A small-scale fusion system based on a reciprocating fusion cycle employing the magneto-kinetic compression of the Field Reversed Configuration (FRC) is particularly well suited for this application. The characteristics of this fusion neutron driver and the potential for transmutation of long-lived nuclear wastes and breeding of fissile nuclear fuel in a blanket are presented.     

Fusion-fission hybrid design with analysis of direct enrichment and non-proliferation features (the SOLASE-H study)

The role of a fusion-fission hybrid in the context of a nuclear economy with and without reprocessing is examined. An inertial confinement fusion driver is assumed and a consistent set of reactor parameters are developed. The form of the driver is not critical, however, to the general concepts. The use of the hybrid as a fuel factory within a secured fuel production and reprocessing center is considered. Either the hybrid or a low power fission reactor can be used to mildly irradiate fuel prior to shipment to offsite reactors thereby rendering the fuel resistant to diversion. A simplified economic analysis indicates a hybrid providing fuel to 10 fission reactors of equal thermal power is insensitive to the recirculating power fraction provided reprocessing is permitted. If reprocessing is not allowed, the hybrid can be used to directly enrich light water reactor fuel bundles fabricated initially from fertile fuel (either ThO₂ or ²³⁸UO₂). A detailed neutronic analysis indicates such direct enrichment is feasible but the support ratio for ²³³U or ²³⁹Pu production is only 2, making such an approach highly sensitive to the hybrid cost. The hybrid would have to produce considerable net power for economic feasibility in this case. Inertial confinement fusion performance requirements for hybrid application are also examined and an integrated design, SOLASE-H, is described based upon the direct enrichment concept.     

Physics principles to achieve comparable fission power from fertile and fissile rods of the conceptual ATBR/FTBR reactors

Loading of seedless fertile rods has been used as the central principle to maximize fertile to fissile conversion in the two thorium breeder reactor concepts, viz. ATBR and FTBR [Jagannathan, V., Pal, Usha, Karthikeyan, R., Ganesan, S., Jain, R.P., Kamat, S.U., 2001. ATBR – a thorium breeder reactor concept for an early induction of thorium in an enriched uranium reactor. Nuclear Technology 133, 1–32; Jagannathan, V., Pal, Usha, Karthikeyan, R., Raj Devesh, Srivastava, Argala, Ahmad Khan, Suhail, 2007. Reactor physics ideas to design novel reactors with faster fissile growth. In: Paper accepted for oral presentation in ‘ICENES 2007 – 13th International Conference on Emerging Nuclear Energy Systems, 3--8 June 2007, Istanbul, Turkey]. At fresh state the seedless thoria rods will produce practically no fission power, or nearly thousand times less fission rate compared to the seed fuel rods. Hence it is conceived that the fuel assembly would be constituted by assembling the fresh seed rods with one fuel cycle irradiated fertile thoria rods. Even in this state there is a wide disparity between the fissile content of these rods. By judicious choice of the rod dimensions and their relative locations, a degree of balance in the fission rate is achieved in the fresh state of seeded rods. Remarkably as the burnup proceeds the initially seedless fertile rods have a continuous growth of fissile content up to an asymptotic value for a given spectrum and the fissile content in seeded rods monotonically decreases. If the discharge burnup is sufficiently large by design, it is seen that the power share of the initially seedless fertile rods can even exceed that of the seed fuel rods. The physics principles of achieving this characteristic are presented in this paper.     

On the proliferation issues of a fusion fission fuel factory using a molten salt

The fusion fission fuel factory (FFFF) is a hybrid fusion fission reactor using a neutron source, which is in this case taken similar to the source of the Power Plant Conceptual Study - Water Cooled Lithium Lead (PPCS-A) design, for fissile material production instead of tritium self-sufficiency. As breeding blanket the first wall of the ITER design is attached to a molten salt zone, in which ThF₄ and UF₄ solute salts are transported by a LiF-BeF₂ solvent salt. For this blanket design, the fissile material is assessed in quantity and quality for both the Th-U and the U-Pu fuel cycle.The transport of the initial D-T fusion neutrons and the reaction rates in this breeding blanket are simulated with the Monte Carlo code MCNP4c2. The isotopic evolution of the actinides is calculated with the burn-up code ORIGEN-S.For the Th-U cycle the bred material output remains below 10 g/h with a ²³²U impurity level of 30 ppm, while for the U-Pu cycle supergrade material is produced at a rate up to 100 g/h.     

Design of coated fuel particles for a hybrid fusion-fission system

Designs have been developed for coated ThO₂ fuel particles to be used in a hybrid fusion-fission system that could be operated without reprocessing. The fresh fertile fuel particle would first be cycled through the blanket of a fusion reactor to breed ²³³U, which would then be ‘burned’ in a thermal fission reactor. The depleted fuel would then be refreshed in a second pass through the fusion reactor, and the process above repeated as many times as feasible. Designs of coated particles for up to three cycles through the hybrid system of reactors have been developed. The outer structural layer for these particles is made from vapor-deposited silicon carbide, because of its remarkable dimensional stability under fast neutron irradiation, and an inner layer of porous pyrocarbon is used to accommodate the buildup of gaseous reaction products inside the particle. The production of gaseous emission products from the interaction of high-energy fusion neutrons with coating materials and with the oxygen in the kernel contributes significantly to pressure vessel stresses in these coatings, whereas gaseous fission products alone dominate in conventional thermal reactors. The most stringent design for the three-cycle particle is identical in fuel loading to the reference fertile particle for an HTGR, which would constitute an ideal hybrid partner for the fusion reactor. Consideration is also given to coated-particle designs for the containment of the bred tritium used to fuel the D-T fusion reactor.     

Prospects for fission gas extraction from in-reactor UO2 nuclear fuel elements

Due to the many problems encountered in the design of fuel rods for the safe operation of commercial nuclear reactors, caused by the fission gases generated by the fission of fissile material, it was considered opportune to make a theoretical analysis of the feasibility of extraction of fission gases from the fuel rod while in operation.This analysis in the steady state of a Zircaloy-2 sheathed fuel rod containing UO₂ as a fuel, with a 2 mm (2.7 vol.%) diameter porous graphite cylinder inserted in the centre, has demonstrated that a total volume of fission gases (xenon, krypton, and iodine) of about 1.1 × 10⁻⁶ cm³/s (at STP) can be extracted from the fuel rod at a controlled rate, determined by the inherent property of fission gas migration towards the centre of the fuel rod from its place of formation. In this analysis, the fuel rod was assumed to be subjected to irradiation in a reactor the size of a Bruce “A” reactor, operating at 3000 megawatts thermal power. The extracted volume of gas was calculated on a 900 h cycle after the first 90 h of reactor operation had elapsed.     

Neutronics Calculations of Performance Enhancement of the Lithium-Uranium Blanket of a Tokamak Reactor

Atomic Energy - The possibility of using a tokamak as a DT-fusion reactor source of fast neutrons for irradiating waste uranium in order to produce energy and fissile fuel (239Pu) for fission...     

Neutronic analysis of a thorium fusion breeder with enhanced protection against nuclear weapon proliferation

The fissile breeding capability of a (D,T) fusion-fission (hybrid) reactor fueled with thorium is analyzed to provide nuclear fuel for light water reactors (LWRs). Three different fertile material compositions are investigated for fissile fuel breeding: (1) ThO₂; (2) ThO₂ denaturated with 10% natural-UO₂ and (3) ThO₂ denaturated with 10% LWR spent fuel. Two different coolants (pressurized helium and Flibe ‘Li₂BeF₄’) are selected for the nuclear heat transfer out of the fissile fuel breeding zone. Depending on the type of the coolant in the fission zone, fusion power plant operation periods between 30 and 48 months are evaluated to achieve a fissile fuel enrichment quality between 3 and 4%, under a first-wall fusion neutron energy load of 5 MW/m² and a plant factor of 75%. Flibe coolant is superior to helium with regard to fissile fuel breeding. During a plant operation over four years, enrichment grades between 3.0 and 5.8% are calculated for different fertile fuel and coolant compositions. Fusion breeder with ThO₂ produces weapon grade ²³³U. The denaturation of the ²³³U fuel is realized with a homogenous mixture of 90% ThO₂ with 10% natural-UO₂ as well as with 10% LWR spent nuclear fuel. The homogenous mixture of 90% ThO₂ with 10% natural-UO₂ can successfully denaturate ²³³U with ²³⁸U. The uranium component of the mixture remains denaturated over the entire plant operation period of 48 months. However, at the early stages of plant operation, the generated plutonium component is of weapon grade quality. The plutonium component can be denaturated after a plant operation period of 24 and 30 months in Flibe cooled and helium cooled blankets, respectively. On the other hand, the homogenous mixture of 90% ThO₂ with 10% LWR spent nuclear fuel remains non-prolific over the entire period for both, uranium and plutonium components. This is an important factor with regard to international safeguarding.     

The fusion breeder

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U₃O₈). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-path-item for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices.     

Computational analysis of neutronic effects of ThO2 rods loaded in CANDU 6 fuel assemblies

Thorium as a suitable fertile with higher natural resources in comparison with uranium resources has been remarkably considered by different nuclear energy user countries in the last decades. Its prominent features such as suitable possibility for power flattening of a nuclear reactor, applicable breeder blanket to produce~(233)U fissile as well as neutron leakage prevention from a nuclear core has caused its application as power flatter, breeder material or other aimed utilizations be evaluated by the researches. In the present study, neutronics of a modeled CANDU 6loaded with Th O_2 and UO_2fuel rods have been computationally studied. The study aimed at reprocessing of burned Th O_2 seeds at CANDU 6 reactor to recover the total produced uranium, which is to be going under another compound fuel cycle. The obtained results showed all the core reactivity coefficients are sufficiently negative. The modeled core 949 GWd burn-up concluding in 99.99 %depletion of~(235)U initial loads. 18.38 kg of~(233) U was produced in the burnt Th O_2 fuel after 1-year burn-up time. In addition, 31.84 kg of~(239) Pu was produced in the UO_2 spent fuel rods after the burn-up time. After a proposed cooling time, about 50.01 kg of~(233)U will be available in the spent Th O_2 fuel.     

Uncertainty quantification of neutronic parameters of light water reactor fuel cells with JENDL-4.0 covariance data

Neutronic parameter uncertainty induced by nuclear data uncertainty is quantified for several light water reactor fuel cells composed of different combinations of fissile/fertile nuclides. The covariance data given in JENDL-4.0 are used as the nuclear data uncertainty, and uncertainty propagation calculations are carried out using sensitivity coefficients calculated with the generalized perturbation theory for burnup-related neutronic parameters.

It is found that main contributors of nuclear data uncertainty to the neutronic parameter uncertainty are the uranium-238 capture cross section in a uranium-oxide fuel cell, and the plutonium-240 and plutonium-241 capture cross sections and fission spectrum of fissile plutonium isotopes in a uranium–plutonium mixed-oxide fuel cell. It is also found that thorium-232 capture cross section uncertainty is a dominant source of neutronic parameter uncertainty in thorium–uranium and thorium–plutonium mixed-oxide fuel cells. It should be emphasized that precise and detail information of component-wise uncertainties can be obtained by virtue of the adjoint-based sensitivity calculation methodology. Furthermore, cross-correlations are evaluated for each fuel cell, and strong correlations among the same parameters at the beginning of cycle and at the end of cycle and among different parameters are observed.     

Optimum utilization of nuclear fuel with gas and vapor core reactors

Gas and Vapor Core Reactors (G/VCR) are externally reflected and moderated nuclear energy systems fueled by stable uranium compound in gaseous or vapor phase. In G/VCR systems the functions of fuel and coolant are combined and the reactor outlet temperature is not constrained by solid fuel-cladding temperature limitations. G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Furthermore, G/VCR systems feature a low inventory and fully integrated fuel cycle with exceptional sustainability and safety characteristics. With respect to fuel utilization, there is practically no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from almost solely a burner to a breeder is achievable. The continuous recycling of fuel in G/VCR systems allows for continuous burning and transmutation of actinides without removing and reprocessing of the fuel. The only waste product at the backend of the gas core reactors' fuel cycle is fission fragments that are continuously separated from the fuel. As a result the G/VCR systems do not require spent fuel storage or reprocessing.

G/VCR systems also feature outstanding proliferation resistance characteristics and minimum vulnerability to external threats. Even for comparable spectral characteristic, gas core reactors produce fissile plutonium two orders of magnitude less than Light Water Reactors (LWRs). In addition, the continuous transmutation and burning of actinides further reduces the quality of the fissile plutonium inventory. The low fuel inventory (about two orders of magnitude lower than LWRs for the same power generation level) combined with continuous burning of actinides, significantly reduces the need for emergency planning and the vulnerability to external threats. Low fuel inventory, low fuel heat content, and online separation of fission fragments are among the key constituent safety features of G/VCR systems.     

Spent fuel reprocessing: A vital link in Indian nuclear power program

The success of the three stage Indian nuclear energy program is inter-linked with the establishment of an efficient closed fuel cycle approach with recycling of both fissile and fertile components of the spent fuel to appropriate reactor systems. The Indian reprocessing journey was started way back in 1964 with the commissioning of a plant based on PUREX technology to reprocess aluminum clad natural uranium spent fuel from the research reactor CIRUS. After achieving the basic skills, a power reactor reprocessing facility was built to reprocess spent fuel from power reactors. Adequate design and operating experience was gained from these two plants for mastering the reprocessing technology. The first plant, being the maiden venture, based on indigenous technology had to undergo many modifications during its operation and finally needed refurbishment for continued operation. Decommissioning and decontamination of this plant was carried out meticulously to allow unrestricted access to the cells for fresh installation. A third plant was built for power reactor spent fuel reprocessing to serve as a design standard for future plants with the involvement of industry. Over the years, spent fuel reprocessing based on PUREX technology has reached a matured status and can be safely deployed to meet the additional reprocessing requirements to cater to the expanding nuclear energy program. Side by side with the developments in the spent natural uranium fuel reprocessing, irradiated thoria reprocessing is also perused to develop THOREX into a robust process. The additional challenges in this domain are being addressed to evolve appropriate technological solutions. Advancements in the field of science and technology are being absorbed to meet the challenges of higher recovery combined with reduced exposure and environmental discharges.     

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.     

Studies on the Feasibility of the LWRs Waste-Thorium In-Core Fuel Cycle in the Gas Turbine-Modular Helium Reactor

The capability to operate on LWRs waste constitutes one of the major benefits of the Gas Turbine-Modular Helium Reactor; in this paper, it has been evaluated the possibility to incinerate the LWRs waste and to simultaneously breed fissile ²³³U by fertile thorium. Since a mixture of pure ²³⁹Pu-thorium has shown a quite poor neutron economy, the LWRs waste-thorium fuel performance has been also tested when plutonium and thorium are allocated in different TRISO particles. More precisely, when fissile and fertile actinides share the same TRISO kernel, the resonance at 0.29 eV of the fission and capture microscopic cross sections of ²³⁹Pu diminishes also the absorption rate of fertile ²³²Th and thus it degrades the breeding process. Consequently, in the present studies, two different types of fuel have been utilized: the Driver Fuel, made of LWRs waste, and the Transmutation Fuel, made of fertile thorium. Since, in the thermal neutron energy range, the microscopic capture cross section of ²³²Th is about 80-100 times smaller than the fission one of ²³⁹Pu, setting thorium in particles with a large kernel and LWRs waste in particles with a small one makes the volume integrated reaction rates better equilibrated. At the light of the above consideration, which drives to load as much thorium as possible, for the Transmutation Fuel they have been selected the JAERI TRISO particles packed 40%; whereas, for the Driver Fuel they have been tested different packing fractions and kernel radii. Since no configuration allowed the reactor to work, the above procedure has been repeated when fertile particles are packed 20%; the latter choice permits over one year of operation, but the build up of ²³³U represents only a small fraction of the depleted ²³⁹Pu. Finally, the previous configuration has been also investigated when the fertile and fissile fuels share the same kernel or when the fertile fuel axially alternates with the fissile one.     

A novel concept of QUADRISO particles – Part III: Applications to the plutonium–thorium fuel cycle

In the present study, a plutonium–thorium fuel cycle is investigated including the ²³³U production and utilization. A prismatic thermal High Temperature Gas Reactor (HTGR) and the novel concept of quadruple isotropic (QUADRISO) coated particles, designed at the Argonne National Laboratory, have been used for the study. In absorbing QUADRISO particles, a burnable poison layer surrounds the central fuel kernel to flatten the reactivity curve as a function of time. At the beginning of life, the fuel in the QUADRISO particles is hidden from neutrons, since they get absorbed in the burnable poison before they reach the fuel kernel. Only when the burnable poison depletes, neutrons start streaming into the fuel kernel inducing fission reactions and compensating the fuel depletion of ordinary TRISO particles. In fertile QUADRISO particles, the absorber layer is replaced by natural thorium with the purpose of flattening the excess of reactivity by the thorium resonances and producing ²³³U. The above configuration has been compared with a configuration where fissile (neptunium–plutonium oxide from Light Water Reactors irradiated fuel) and fertile (natural thorium oxide) fuels are homogeneously mixed in the kernel of ordinary TRISO particles. For the ²³³U utilization, the core has been equipped with europium oxide absorbing QUADRISO particles.     

Potential of CANDLE reactor on sustainable development and strengthened proliferation resistance

CANDLE reactor generates energy by using only natural or depleted uranium as make up fuel and achieves about 40% burn up without fuel recycling (Sekimoto, H., Ryu, K., Yoshimura, Y., 2001. CANDLE: the new burnup strategy. Nucl. Sci. Eng. 139 (3), 306–317). These distinctive characteristics eliminate the necessity of both enrichment and reprocessing processes that are recognized as essentially inevitable parts in the conventional nuclear energy concept.

This paper describes that the potential performance of CANDLE reactor to meet a projected energy demand growth of the world with stabilizing carbon dioxide concentration in the atmosphere and simultaneously minimize the risk of nuclear material proliferation.

A type of CANDLE reactor with moderate initial fissile inventory is feasible to be deployed with prompt enough introduction pace to satisfy the worldwide energy growth and limit the carbon dioxide concentration to about 550 ppm in the next century. Ultimately high proliferation resistance performance as a fission energy system is found for the CANDLE system due to the elimination of most vulnerable processes in the conventional fuel cycle.     

The tokamak hybrid reactor

At a time when the potential benefits of various energy options are being seriously evaluated in many countries throughout the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the Commercial Tokamak Hybrid Reactor (CTHR) [1]. This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of ‘client’ Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode.