Fusion-fission Hybrid System Research Articles

An approach to the experimental validation of the fission multiplying blanket of hybrid fusion fission systems

Fusion Fission Hybrid Systems (FFHS) could have a potential role in the management of fission reactor wastes, with (in principle) some advantage over the comparable Accelerator Driven systems, devoted to the same objective. The validation of the concept poses major challenges in the area of fusion “source” development. However, the physics of the multiplying and transmutation blanket has also to be carefully understood and validated. Simple experiments can be envisaged for that purpose. Some of these experiments have already been performed in a different research framework (i.e. validation of the ADS concepts) and could constitute a very first database. Moreover, a new series of similar experiments could be planned.

System Studies on the Fusion-Fission Hybrid Systems and Its Fuel Cycle

This paper is devoted to applications of fusion-fission hybrid systems (FFHS) as a powerful neutron source implementing transmutation of minor actinides (MA: Np, Am, Cm) extracted from the spent nuclear fuel (SNF) of nuclear reactors. Calculations which simulated nuclide kinetics for the metallic fuel containing MA and neutron transport were performed for particular facilities. Three FFHS with fusion power equal to 40 MW are considered in this study: demo, pilot-industrial and industrial reactors. In addition, needs for a fleet of such reactors are assessed as well as future FFHSs’ impact on Russian Nuclear Power System. A system analysis of nuclear energy development in Russia was also performed with the participation of the FFHSs, with the help of the model created at AO “Proryv”. The quantity of MA that would be produced and transmuted in this scenario is estimated. This research shows that by the means of only one hybrid facility it is possible to reduce by 2130 the mass of MA in the Russian power system by about 28%. In the case of the absence of partitioning and transmutation of MA from SNF, 287 t of MA will accumulate in the Russian power system by 2130.

Fusion–Fission Hybrid Systems: Yesterday, Today, and Tomorrow

Fundamentally new characteristics and parameters can be achieved in systems for energy production and special applications by combining nuclear fusion and fission reactions in a single design. In the first decades of the atomic era, fusion-fission hybrid systems (FFHSs) were developed for military applications. The civilian use of hybrid systems in the energy sector, which was also foreseen by the creators of nuclear explosives, has proved to be much more difficult. This paper discusses the development of FFHSs: the initial stage from 1950 to 2000, the current stage from 2000 to 2020, and the long-term targets for 2020 to 2130.

Tritium Breeder Layer Evaluation of Fusion-Fission Hybrid System

Lead-bismuth eutectic is used as a coolant for the fusion-fission hybrid system (FFS) based on a tokamak that enhances the transmutation of transuranic nuclides. However, this coolant does not produce enough tritium to supply the fusion reactions of the system. Therefore, the aim of this work is to evaluate the insertion of tritium breeder layers (TBLs) on the FFS to enhance tritium production. The analyzed materials for tritium production were beryllium, boron, and lithium alloys. The results indicate the most suitable material for tritium production depends on the TBL location. The results also indicate that there is a strong dependency on the position of the TBL affecting the neutronic parameters and nuclide transmutation such as criticality and fuel depletion. The reaction rates for tritium production and fuel composition after a fuel burnup were analyzed using the Monte Carlo N-Particle 5 (MCNP5) and MONTEBURNS codes.

On the feasibility of fusion–fission hybrid based on deuterium fuelled tokamak

The efficiency of fast neutrons generation in a deuterium tokamak plasma is analyzed. Tokamak parameters were estimated by relations of ITER Physics Basis. The parameters of D–D cycle are fundamentally different from the case of the D–T cycle with low tritium content. In an ITER scale tokamak with an aspect ratio A = 3, the gain and power are very low. In a D–D fuelled tokamak with A = 2 the power gain reaches Q ∼ 0.3. The efficiency of a fusion–fission hybrid system with the production of tritium in the D–D reaction was investigated. Two types of the tritium cycle are considered. In the first case, all tritium is burned in a catalyzed D–D cycle in a tokamak with A = 2. In the second case, some portion of tritium is burned in D–D based system, and some portion is used as a component of D–T fuel for a tokamak with A = 3 of relatively smaller size and greater efficiency.

Progress in the design of the DEMO-FNS hybrid facility

Further development of a fusion–fission hybrid facility based on a superconducting tokamak, DEMO-FNS, continues in Russia for integrated commissioning of steady-state and nuclear fusion technologies at a power level up to 40 MW fusion and 400 MW fission reactions. This facility is considered as the main source of technological and nuclear scientific information in the RF National program on controlled fusion and plasma technologies that is currently being developed and submitted to the authorities for approval. In this work the progress over the last two years is reported in core plasma modelling, design and integration of enabling systems, including the first wall, divertor, neutron beam heating and current drive, as well as the choice of materials. Use of supercritical CO2 coolant was reviewed from an activation point of view. Interaction of fusion–fission hybrid systems with the nuclear fuel cycle of Russia’s nuclear power industry was shown. Сonstruction of the device is expected in 2033.

Neutronics performance analysis on neutron consumption in a Fusion-Fission Hybrid System for tritium breeding

Comparison evaluation of system performance for Fusion-Fission Hybrid System as a tritium supply facility to perform D-T fusion researches was conducted. Performances for neutron consumption as well as k-eff variation, energy multiplication, tritium breeding and waste transmutation were investigated. All neutronics calculations were conducted by MCNPX2.6.0 and MCNP6.1 with ENDF/b-VII.0 neutron cross-section library.Analysis results for tritium breeding modules show that tritium with module using water or FLiNaBe coolant is effectively produced due to softening effect in spite of low 6Li enrichment. When designing hybrid system for tritium breeding, design of the fuel zone between tritium breeding zones is beneficial for tritium breeding. In this case, 7 kg of tritium is more produced without degradation of transmutation compared to the design of fuel zone in front of the tritium breeding zone. In the comparison evaluation of performance on coolants options; PbBi, LiPb, Na and FLiNaBe, tritium with Li containing coolant is produced in the fuel zone, however amount of produced tritium in the tritium breeding zone is reduced due to low neutron economy. Performance of tritium breeding with softening flux is degraded, since neutron economy is reduced by increasing the capture reaction.

Neutronic investigation of waste transmutation option without partitioning and transmutation in a fusion-fission hybrid system

Abstract A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial keff level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially 241Am is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

Nuclear Problems of Thermonuclear Power Generation

The current status of the nuclear problems of the thermonuclear fusion research program and, in particular, tritium, is discussed. It is noted that thermonuclear power generation without a uranium or thorium blanket is problematic; the key nuclear problems of the fusion–fission hybrid system remain unsolved. It is proposed that an integrated strategic analysis be made of the thermonuclear research program and the realistic possibilities of its application in nuclear power-engineering.

Transmutation considerations of LWR and RBMK spent nuclear fuel by the fusion–fission hybrid system

The performance of the fusion–fission hybrid system based on the molten salt (flibe) blanket, driven by a plasma based fusion device, was analyzed by comparing transmutation scenarios of actinides extracted from the LWR (Sweden) and RBMK (Lithuania) spent nuclear fuel in the scope of the EURATOM project BRILLIANT. The IAEA nuclear fuel cycle simulation system (NFCSS) has been applied for the estimation of the approximate amount of heavy metals of the spent nuclear fuel in Sweden reactors and the SCALE 6 code package has been used for the determination of the RBMK-1500 spent nuclear fuel composition. The total amount of trans-uranium elements has been estimated in both countries by 2015. Major parameters of the hybrid system performance (e.g., kscr, keff, Φn(E), equilibrium conditions, etc.) have been investigated for LWR and RBMK trans-uranium transmutation cases. Detailed burn-up calculations with continuous feeding to replenish the incinerated trans-uranium material and partial treatment of fission products were done using the Monteburns (MCNP + ORIGEN) code system. About 1.1 tons of spent fuel trans-uranium elements could be burned annually with an output of the 3 GWth fission power, but the equilibrium stage is reached differently depending on the initial trans-uranium composition. The radiotoxicity of the remaining LWR and RBMK transmuted waste after the hybrid system operation time has been estimated.

Tokamak DEMO-FNS: Concept of magnet system and vacuum chamber

The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion—fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet system and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k 95 = 2. The fusion power is P FUS = 40 MW. The toroidal magnetic field on the plasma-filament axis is B t0 = 5 T. The plasma current is I p = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb3Sn, NbTi and Nb3Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron—water radiation shield (70% of stainless steel and 30% of water).

Thorium Fuel Cycles with Externally Driven Systems

Externally driven subcritical systems are closely associated with thorium, partially because thorium has no naturally occurring fissile isotopes. Both accelerator-driven systems (ADSs) and fusion-driven systems have been proposed. This paper highlights key literature related to the use of thorium in externally driven systems (EDSs) and builds upon this foundation to identify potential roles for EDSs in thorium fuel cycles. In fuel cycles with natural thorium feed and no enrichment, the potential roles are (1) a once-through breed-and-burn fuel cycle and (2) a fissile breeder (mainly 233U) to support a fleet of critical reactors. If enriched uranium is used in the fuel cycle in addition to thorium, EDSs may be used to burn transuranic material.These fuel cycles were evaluated in the recently completed U.S. Department of Energy Evaluation and Screening of nuclear fuel cycle options relative to the current once-through commercial nuclear fuel cycle in the United States. The evaluation was performed with respect to nine specified high-level criteria, such as waste management and resource utilization. Each of these fuel cycles presents significant potential benefits per unit energy generation compared to the present once-through uranium fuel cycle. A parametric study indicates that fusion-fission–hybrid systems perform better than ADSs in some missions due to a higher neutron source relative to the energy required to produce it. However, both potential externally driven technology choices face significant development and deployment challenges. In addition, there are significant challenges associated with the use of thorium fuel and with the transition from a uranium-based fuel cycle to a thorium-based fuel cycle.

First wall material damage induced by fusion-fission neutron environment

Neutronic performance and inventory analyses were conducted to quantify the damage and gas production rates in candidate materials when used in a fusion-fission hybrid system first wall (FW). The structural materials considered are austenitic SS, Cu-alloy and V- alloys. Plasma facing materials included Be, and CFC composite and W. It is shown that the highest damage rates and gas particles production in materials are experienced within the FW region of a hybrid similar to a pure fusion system. They are greatly influenced by a combined neutron energy spectrum formed by the two-component fusion-fission neutron source in front of the FW and in a subcritical fission blanket behind. These characteristics are non-linear functions of the fission neutron source intensity. Atomic displacement damage production rate in the FW materials of a subcritical system (at the safe subcriticality limit of ∼0.95 and the neutron multiplication factor of ∼20) is almost ∼2 times higher compared to the values distinctive for a pure fusion system at the same 14MeV neutron FW loading. Both hydrogen (H) and helium (He) gas production rates are practically on the same level except of about ∼4–5 times higher He-production in austenitic and reduced activation ferritic martensitic steels. A proper simulation of the damage environment in hybrid systems is required to evaluate the expected material performance and the structural component residence times.

Fusion–Fission Hybrid Systems for Transmutation

Spent fuel reprocessed by UREX+ technique and spiked with thorium was submitted to a neutron irradiation in two different fusion–fission systems. One based on magnetic confinement and the other one on inertial confinement. The goal is to compare and find the better fusion system to induce transmutation with increased its efficiency. Results indicate that the fusion system based on a Tokamak has greater transmutation values.

Progress in Mirror-Based Fusion Neutron Source Development.

The Budker Institute of Nuclear Physics in worldwide collaboration has developed a project of a 14 MeV neutron source for fusion material studies and other applications. The projected neutron source of the plasma type is based on the gas dynamic trap (GDT), which is a special magnetic mirror system for plasma confinement. Essential progress in plasma parameters has been achieved in recent experiments at the GDT facility in the Budker Institute, which is a hydrogen (deuterium) prototype of the source. Stable confinement of hot-ion plasmas with the relative pressure exceeding 0.5 was demonstrated. The electron temperature was increased up to 0.9 keV in the regime with additional electron cyclotron resonance heating (ECRH) of a moderate power. These parameters are the record for axisymmetric open mirror traps. These achievements elevate the projects of a GDT-based neutron source on a higher level of competitive ability and make it possible to construct a source with parameters suitable for materials testing today. The paper presents the progress in experimental studies and numerical simulations of the mirror-based fusion neutron source and its possible applications including a fusion material test facility and a fusion-fission hybrid system.

Parameters of a Fusion Neutron Source Based on the Recent GDT Experimental Data and Possible Applications

Substantial progress in experimental results was demonstrated over the last three years at the GDT facility (a hydrogen prototype of a fusion neutron source) in the Budker Institute: the electron temperature has been increased up to 0.6 keV, and the relative plasma pressure β has exceeded 0.5 in a quasi-stationary regime. These parameters are records for axisymmetric open mirror traps.The first part of this paper presents the results of numerical simulations for a moderate fusion neutron source based on the achieved GDT experimental data. The second part of the paper is focused on the latest numerical studies of a fusion-fission hybrid system with a mirror-based neutron source.

Development of DEMO-FNS tokamak for fusion and hybrid technologies

The history of fusion–fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ∼ 1, fusion power of ∼40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

LWR spent fuel transmutation with fusion-fission hybrid reactors

In this paper the transmutation of light water reactors (LWR) spent fuel is analyzed. The system used for this study is the fusion-fission transmutation system (FFTS). It uses a high energy neutron source produced with deuterium-tritium fusion reactions, located in the center of the system, which is surrounded by a fission region composed of nuclear fuel where the fissions take place. In this study, the fuel of the fission region is obtained from the recycling of LWR spent fuel. The MCNPX Monte Carlo code was used to setup a model of the FFTS. Two fuel types were analyzed for the fissile region: the mixed oxide fuel (MOX), and the inert matrix fuel (IMF). Results show that in the case of the MOX fuel, an important Pu-239 breeding is achieved, which can be interesting from the point of view of maximal uranium utilization. On the contrary, in the case of the IMF fuel, high consumption of Pu-239 and Pu-241 is observed, which can be interesting from the point of view of non-proliferation issues. A combination of MOX and IMF fuels was also studied, which shows that the equilibrium of actinides production and consumption can be achieved. These results demonstrate the versatility of the fusion-fission hybrid systems for the transmutation of LWR spent fuel.

Axisymmetric Magnetic Mirror Fusion-Fission Hybrid

Early application of the simple axisymmetric mirror, requiring intermediate performance between a neutron source for materials testing Q=Pfusion/Pinput ~0.05 and pure fusion Q>10, are the hybrid applications. The Axisymmetric Mirror has attractive features as a driver for a fusion-fission hybrid system: geometrical simplicity, as well as the typical mirror features of inherently steady-state operation, and natural divertors in the form of end tanks. This level of physics performance has the virtue of being low risk with only modest R&D needed; and its simplicity promises economy advantages. Operation at Q~1 allows for relatively low electron temperatures, in the range of 3 keV, for the DT injection energy ~ 80 keV from existing positive ion neutral beams designed for steady state. A simple mirror with the plasma diameter of 1 m and mirror-to-mirror length of 40 m is discussed. Simple circular steady state superconducting coils are based on 15 T technology development of the ITER central solenoid. Three groups of physics issues are presented: axial heat loss, MHD stability, and microstability of sloshing ions.Burning fission reactor wastes by fissioning transuranics in the hybrid will multiply fusion’s neutron energy by a factor of ~10 or more and diminish the Q needed to overcome the cost of recirculating power for good economics to less than 2 and for minor actinides with multiplication over 50 to Q~0.2. Hybrids that obtain revenues from sale of both electricity and production of fissile fuel with fissioning blankets might need Q<2 while suppressing fissioning might be the most economical application of fusion but will require Q>4.

Numerical Model of the Fusion-Fission Hybrid System Based on Gas Dynamic Trap for Transmutation of Radioactive Wastes

The paper presents a 3D numerical model of the neutron source for the transmutation of long-lived radioactive waste in spent nuclear fuel. The projected plasma type neutron source is based on the Gas Dynamic Trap (GDT) which is a special magnetic mirror system for the plasma confinement. A new improved version of the GDT type fusion neutron source is numerically simulated by use different numerical methods. New physical phenomena such as a vortex confinement, improved axial confinement, low radial transport, high β etc. were included in these simulations. The experimental and theoretical foundations of these phenomena were obtained in the GDT-U experimental facility in the Budker Institute. In result the proposed neutron source has two n-zones of 2 m length with a neutron power of 1.6 MW/m and a neutron production rate up to 1.5x1018 n/s each. This source can be used for application to a fusion driven system for the burning of MA in spent nuclear fuel.