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The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion–fission reactor.

The stellarator plasma device has been widely studied as one of the candidate solutions paralleling the ITER project, and its coupling with a fission blanket can bring benefits promoting the develo...

The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors. Keywords: LWR spent fuel, Thorium, Time dependent neutronic analysis, Fusion-fission hybrid reactor

As an important parameter of neutronics integral experiment, tritium production rate could not only be used for validation of programs and evaluation database, but was also related to tritium self-consistency after the activation of the sub-critical reactor. This parameter was a key component of the sub-critical reactor design. Four blanket benchmark mockups of a fusion-fission hybrid reactor were established utilizing lithium materials, which consisted of depleted uranium/lithium hydride, uranium/lithium hydride/polyethylene, uranium/lithium hydride/graphite/iron, and uranium/Li-Pb alloy/polyethylene. Based on the Cockcroft–Walton neutron generator, the distribution of tritium production rate by D–T neutrons in the four mockups was evaluated by two pairs of different sizes lithium glass detectors. The distribution was then quantitatively assessed using MCNP5 program and ENDF/B-VII database. The tritium production rates of the mockups with lithium hydride significantly deviated from the experimental results; whereas those of the mockup with lithium-lead alloy were agree well with the experimental results.

Barbotage reactors such as airlift reactors (ALR) and bubble column reactors (BCR), due to their two-phase flow systems, were investigated in many research papers. In their basic design variants, they are typically used to lift, mix, and aerate liquids, while, when equipped with additional elements in hybrid variants, their individual properties, i.e., lifting, mixing, and aeration of liquids, can significantly change with the same reactor geometry. The object of this study was to develop a hybrid barbotage reactor in various structural design variants. The structure consisted of a barbotage column of 50 mm in diameter, used to transport a water&ndash, air mixture outside the reactor (so-called external loop). The installation was additionally equipped with a nozzle in order to improve mixture aeration and circulation efficiency. The nozzle was mounted at various heights of the column pump segment. Additionally, the reactor was equipped with s moving bed in two variants (20% and 40% reactor capacity) in order to determine its effect on the mixture aeration and circulation conditions. Based on the measurement results, aeration curves were prepared for various structural design and column packing variants of the reactor. Properties of the two-phase mixture were determined for both parts&mdash, ALR and BCR. Technological and energy parameters of the aeration process were calculated, and the results obtained for the individual structural design variants were compared. It was found that, for the most advantageous design, in terms of aeration efficiency, the aeration nozzle should be placed in the mid-length of the pump segment of the barbotage column, irrespective of the hybrid reactor packing rate with the moving bed. The reactor packing with the moving bed resulted in a decreased mean water velocity in the reactor. For most analyzed structural design variants, the respective packing with the moving bed had no significant effect on aeration efficiency. Only for one structural design variant did the lack of packing significantly improve oxygen levels by as much as approximately 41%.

Up today, two hyper research projects to achieve nuclear fusion energy exist; inertial confinement fusion (ICF) driven by laser, called national ignition facility (NIF) and magnetic confinement fusion the international thermonuclear experimental reactor (ITER) project. In reaching the required temperature and pressure, to ignite nuclear fusion reactor, is technologically complex and economically expensive. Thus, a breakthrough and a short cut, other alternative methods should be considered. Pulsed power ICF driver with repetitive pulse operation, mainly dense plasma focus (DPF) machines for high yield fusion neutrons could be taken as drivers for the fission blanket operation. The setup can be a cost-effective and efficient. In this article, we consider a set of two medium energy sizes DPF to produce simultaneously dense plasma columns, operating as thermonuclear plasma driver, to pierce the pellet target for external nuclear fusion reactions. These DPFs produce sufficient fast neutrons for the fission process in the neutral uranium or thorium and/or weak enriched uranium blanket. The drive systems and the concept for delivering thermonuclear plasma to pellets target in the magnetic free zone of central region will be presented. The feasibility of such fusion–fission hybrid reactor will be discussed.

Fusion-fission hybrid reactor is believed to be a feasible implementation of fusion energy in a short time as it requires less advanced technology in fusion physics and engineering than pure fusion reactor. In this paper, based on the mature technology of pressurized water reactor (PWR) and existing fusion reactor blanket concepts, a new conceptual design scheme of water-cooled pressure tube blanket was proposed for hybrid reactor and both the material selections and the structure design were performed. In the blanket, a trapezoidal U-shaped first wall (FW) was adopted to decrease the gaps between blanket modules. Due to high heat flux from plasma and intense irradiation by high-energy neutrons, it's essential to conduct thermal-mechanical coupling analyses and optimizations of the FW to ensure reasonable temperature and stress distributions on it. In this work, CFX-Workbench coupling method was adopted to perform the simulation. The influences of different FW geometrical configurations and boundary conditions on the temperature and stress distributions were researched in detail to optimize the FW structure. Besides, the thermal and mechanical characteristics of the helium tanks, the cooling channels and the breeding zones were researched in this paper. The calculation results verified the reasonability of the conceptual design scheme preliminarily and could provide a meaningful reference for the further conceptual design, analysis and optimization of the hybrid reactor blanket.

In this paper, an integrated fusion-fission hybrid reactor energy system aiming at utilizing the thorium fuel efficiently with Th-U closed cycle is presented. The ultimate purpose is transition to the all thorium fueled energy system with the bred 233U fuel burning in-site. The system consists of two fusion-fission hybrid reactors. The first reactor is a fast breeder which provides the starting fuel 233U for the second reactor (thermal burner). The bred 233U is burnt in the second reactor with the thorium bearing molten salt fuel blanket. In order to reach the multi-system performance, i.e. high energy multiplication M, tritium self-sufficiency and fissile fuel self-sufficiency, using limiting 233U start fuel, the seed-blanket concept is adopted in the thermal burner design. The thermal burner blanket is divided into two main regions. The front region (seed) plays the dominate role in the energy production, neutron multiplication and 233U breeding. Excess neutrons produced by the front region are then used to drive backward region (blanket) to breed tritium. The Couple 2.0 code developed by Tsinghua University is used to calculate the fuel depletion process of the whole system. The simulated results show, the thermal burner could operate with excellent performance, energy multiplication M (8–10), tritium breeding ratio TBR (>1.05) continuously and only natural thorium fuel is added into the system.

Fusion-fission hybrid reactor takes advantage of the best of both domains, fusion and fission, without needing a full development of fusion reactor, and without critical safety issues of fission reactor. However, decay heat still exists in fission blanket of hybrid rector in shutdown state, which will menace the safety of rector if it cannot be smoothly carried out. To testify the passive safety characteristics of fission blanket, the reactor coolant pump shaft seizure accident in fission blanket of FFHER, a fusion-fission hybrid reactor, is simulated. The model of passive residual heat removal system, as well as reactor coolant system, are built using software SCDAP/RELAP5 4.0. The analysis shows that the simulated system response is reasonable. In the SG domination stage, the peak fuel temperature keeps always below the limiting value. In the PRHR domination stage, three cases are simulated and compared to choose appropriate parameter. It is found that PRHR circulation can be rapidly established and transfer enough heat from blanket. The transition from SG domination to PRHR domination is smooth. From the simulation results, it can be concluded that PRHRS can provide sufficient blanket cooling capability in LOFA.

A simple modification of the fission source spectrum is proposed for the Monte Carlo calculation for externally driven subcritical reactors. Embedded fission source (EFS) approach replaces the fission spectrum with the collective spectrum of the external source and fission working in a small region near to external source only. This model provides eigenvalue search with external source contribution resulting in direct connection with isotope depletion calculation within Monte Carlo simulation. MCNPX 2.6.0 is modified for EFS and tested for three kinds of reactor problems; critical fast reactor driven subcritical reactor, accelerator driven subcritical reactor and fusion-fission hybrid reactor. The effectiveness of the external sources is determined by the extra amount of the TRU isotopes burnt and increase in the core multiplication level. The tests were performed for two subcritical cases; keff ∼ 0.986 and keff ∼ 0.959, with initial TRU loading of 188.8 kg and 183.2 kg, respectively. The EFS approach is validated by simulating a fission source driven system which gave the multiplication factor comparable to the original MCNPX within one standard deviation (32 pcm). The difference in TRU burn-up for both subcritical cases is of the order of 0.1% for each driving neutron source option. The study showed that the EFS can be a simple and a reliable approximation for the external source analysis.

Within the framework of activities on a pilot industrial hybrid reactor (Pilot Hybrid Plant – PHP) development, the DEMO-FNS fusion neutron source is being designed in Russia. Russian fusion and fission community considers it as the main facility developing prospective nuclear technologies, which should update the results of the ITER researches in the field of physics and control of burning plasma. Progress in the development of the plant and simulations of the fusion fuel cycle allows starting the engineering design of basic fuel cycle systems. In this paper, we consider the new prospective of technological solutions for the tokamak fuel cycle. The possibility of their integration and the analysis of the technology readiness level for them in the Russian Federation are addressed as well. The code simulating the hydrogen isotope flows in a tokamak was significantly upgraded for evaluation of the specific technological solutions as well as for combined modeling of the fuel circulation in the core and divertor plasmas allowing for the pumping, processing, and fuel isotope content control in the plasma. The article describes the core and divertor plasma model used for the calculations and shows the D/T fuel throughput values, which should be provided by the fuel injection systems (NBI and pellets). The dependencies of the HFS and LFS pellet injection frequency necessary for the main plasma supply and ELM control are found for different core and divertor plasma parameters. For the DEMO-FNS core plasma, an operating window has been found in terms of fueling control parameters. It was shown that D and T throughputs of 40-50 Pa•m3/s should be provided by the fuel injection systems (NBI and pellets) to maintain the fusion power up to 40 MW. For the DEMO-FNS core plasma, an operating window defined by the particle confinement times and fuel injection throughputs has been found.

In this technical note, an innovative thorium-uranium–fueled fusion-fission hybrid reactor (FFHR) design that employs a dual-coolant system to enhance 233U breeding and is based on a three-dimensional engineering model is presented. The reactor consists of two kinds of modules: a water-cooled, thermal spectrum power generation natural uranium–fueled module and helium-cooled, fast spectrum fissile-breeding natural thorium–fueled modules, which are arranged alternately in the poloidal direction of the blanket. An interesting and important neutronic characteristic of the FFHR is found in this technical note: Energy multiplication is primarily determined by the uranium module parameters and is almost independent of the thorium module parameter. Uranium module design should first consider improving energy production. The 232Th neutron capture rate is primarily determined by the thorium module parameters. The uranium module parameter has almost no influence on the 232Th neutron capture rate in the thori...

An accelerator-driven system (ADS) combined with a subcritical molten salt reactor (MSR) is a type of hybrid reactor originally designed to use Th/U (or U/Pu ) fuel cycles. In most accelerator-driven molten salt reactor (AD-MSR) concepts, the salt material is also used as a target for inducing spallation neutrons. Although a neutron source is an important component in the design of ADS, only a few studies have addressed the effects of the neutron spallation source in the AD-MSR. Incidentally, there is no quantitative study on how much the beam power can be reduced by installing a spallation target in a sodium chloride-based fast reactor. We studied the proton and the neutron source efficiencies of an AD-MSR with chloride fuels by considering an Lead Bismuth Eutectic (LBE) spallation target. This LBE target is found to increase the proton source efficiency significantly. The required beam power for an AD-MSR can be reduced by 33 % and 16 % for NaCl-Th/233U and NaCl-U/Pu fuels, respectively, relativ...

The paper describes the concept of a deuterium-tritium fuel cycle of a steady-state thermonuclear reactor with a fusion power over 10 MW. Parameters of fuel cycle for nuclear facility (JET scale) with different types of fuel mixtures for neutral beam injection system are presented. Optimization of fuel cycle characteristics was aimed at reducing flows and inventory of hydrogen isotopes and tritium in fuel cycle subsystems. The calculations were carried out using computer code TC-FNS to estimate tritium distribution in fusion reactor systems and components of “tritium plant”. The code enables calculations of tritium flows and inventory in the tokamak systems. Calculations of tritium flows and accumulation have been carried out for two different cases of the fuel mixture for neutral beam injection (NBI) system. The amounts of tritium which is required for operation of all fuel cycle systems in two different cases of the fuel mixture for NBI are 0.45 “” kg (D:T = 1:0) and 0.9 kg (D:T = 1:1) respectively.

The Reversed Field Pinch (RFP) configuration looks to be an attractive option for fusion-fission hybrid reactors: the toroidal magnetic systems would be made of copper coils instead of more expensive superconductive magnets; fusion conditions could be reached by ohmic heating only, therefore additional heating systems would not be required; the fission blanket could be located in the most external part of the torus thus facilitating maintenance operations. The paper aims at assessing the potentialities, such as fuel fertilization and/or nuclear waste transmutation and electricity production of a hybrid reactor with a RFP fusion core (R = 6 m, a = 1) whose conceptual design and plasma performances are based on RFX-mod, the largest RFP experiment currently in operation. Fusion conditions can be reached by heating a D-T plasma up to 9.6 keV by ohmic heating, generated by a 20 MA plasma current induced and sustained by flux swing only. The neutron flux (2.1 × 1013 fast neutron/cm2/s) is used to breed tritium in both the inner and outer blanket sections and induce fission reactions in dedicated areas in the external blanket section where Pu + MA (60%)-Zr (40%) rods are located. Both neutronic and safety analyses corroborate the viability of a FFH reactor with a RFP core.

Solar thermochemical gasification of biomass promises a new path for synthetic fuels generation. However, it suffers from critical barriers related to solar energy variability caused by cloud passages and shut off at night. In the present work, we have investigated by simulation the concept of hybridized solar/autothermal gasification reactors. The goal was to demonstrate the operational feasibility of the hybrid process to stabilize the gasification temperature by injecting a controlled amount of oxygen and burning a portion of the feedstock for supplying the deficient process heat. The proposed CFD model was developed by taking into account the coupled momentum, heat and mass transfer and the chemical reactions (pyrolysis, heterogeneous char gasification and gas-phase reaction kinetics) along with the continuous particle injection. It was first validated against our experimental data during allothermal operation at 1200°C, 1300°C and 1400°C. Then, combustion reactions were integrated to the model to predict the hybridized reactor performance and synthesis gas composition. Moreover, first experimental results of the hybrid reactor operation were shared. The study has demonstrated the effectiveness and the feasibility of process hybridization for a directly irradiated solar gasifier.

The combination of biomass and solar energy in a solar-driven thermochemical gasification process is of particular interest to convert solid carbonaceous materials to syngas, consisting chiefly of CO and H2, thereby offering an efficient means of storing intermittent solar energy into carbon neutral solar fuels. The development of solar gasifiers suffers from the inherent issue of solar energy variability caused by cloud passages and shut off at night. The concept of hybrid solar/autothermal reactors is proposed to meet the requirement of continuous operation under fluctuating solar radiation. The hybrid reactor harvests the thermal energy from both solar heat (allothermal) and partial oxy-combustion of the feedstock (autothermal) if the solar resource is not high enough to reach the minimum temperature for the gasification reaction. In this work, for the first time a directly and indirectly irradiated hybrid reactor subjected to real concentrated solar flux was experimentally studied and modelled. Effects of O2 injection on temperature distribution, gas production rates and LHV are discussed. In order to provide a better overview of the reactor thermal, chemical and hydrodynamic behavior, a 3D Computational Fluid dynamics (CFD) model was developed. Results in terms of temperature and velocity profiles, reactive particles trajectories inside the reaction chamber, syngas composition for the allothermal and hybrid model are provided. The study showed that process hybridization is relevant to control the reactor temperature although significantly impairing the syngas yield and quality.

The stellarator-type Fusion-Fission Hybrid Reactor (FFHR) is an alternative conceptual design for the tokamak-based devices in order to realize the stable power output. It can bring the viable energy source for the fusion technology in a foreseeable future, and has been widely studied with a lot of advanced concepts being proposed. However, there are some challenges in the neutronics design of the FFHR blanket for the purpose of obtaining the appropriate Tritium Breeding Ratio (TBR) and blanket energy multiplication factor, since the blanket system has a complicated three-dimensional structure in helical shape, which can not be easily translated into native geometry representation syntax of most Monte Carlo neutron transport codes. In this paper, a liquid fuel, F-Li-Be molten salt, with uranium and thorium dissolved has been proposed for the preliminary design of stellarator-type FFHR blanket. Besides, the CAD-PSFO code [ 19 ] has been employed for the simulation of complex geometry structure in the helical blanket, and the OMCB package [ 22 ] has been used in the burnup analysis of the neutronics features in the molten salt blanket. Moreover, detailed geometry design of the blanket arrangement and thickness have been studied in order to meet the requirement of energy multiplication factor in the stellarator-type FFHR. Also, the composition and concentration of the fuel material and 6Li in molten salt have been discussed for the purpose of tritium self-sufficiency with steady energy supply in the whole lifetime.

Fusion - fission hybrid reactor consists of two main parts. The first part is the Dense Plasma Focus (DPF) where the fusion process is performed and the second part is the blankets which surround the DPF and the fission process is done in the blankets. The most of the fusion neutrons, near the fusion source, are directed upward, so the blanket over the DPF is designed in forms cylindrical, spherical and frustum cone of volumes 183,347 cm3, 136,378 cm3 and 58,223.2 cm3 respectively at the same effective multiplication factor. Then the thermal neutron diffusion equation, buckling and neutron parameters are studied. After selecting the first blanket, the second blanket is designed in forms cylindrical, spherical and frustum cone of volumes 4.500 × 106 cm3, 3.689 × 106 cm3 and 6.441 × 105 cm3 respectively. So the frustum cone is chosen for the first and second blanket. The neutron return probability in the frustum cone from the second blanket to the first blanket is low, so the first blanket's effective multiplication factor does not change and the reactor's total effective multiplication factor is stable. Also, MCNPX code has been used to checkout and select neutron multiplier and reflector cells for the most neutron yield.

The stellarator plasmas devices have some advantages in steady state operation with less magneto hydrodynamic activities, and the development of stellarator-type FFHR (Fusion-Fission Hybird Reactor) can provide viable energy in the foreseeable future with less requirements to the plasmas technology compared with the pure fusion system. However, the neutronics design of the stellarator-type FFHR has some challenges and difficulties, since the construction of helical structures with large number of high order spline surfaces are the extremely complicated process, and the physical parameters of the complex model have to be changed frequently in the primary design stage, which make the corresponding manual input simulation work into the time consuming and error prone tasks. In this context, the CAD optimization method has been used in converting the dense mesh oriented CAD structure to the low poly style, and the CAD-PSFO code has been employed for the modeling transformation process in Monte Carlo calculation. The thorium resource has been selected as the breeding material for the stellarator-type FFHR, and the polonium and natural abundance of uranium have been used as the seed fuel for the initial loading in order to get the high energy multiplication factor. The OMCB (OpenMC based Burnup code) has been introduced to study the burnup features, and stellarator-type FFHR has been designed as the multi-purposes facility that can keep tritium self-sufficiency and get the high breeding ratio with steady energy output.

A comprehensive comparison neutronics calculation has been performed to investigate the effect of different reflectors on the tritium production and transmutation of minor actinides in a fusion hybrid reactor. Fusion-Driven Subcritical Spent Fuel Burner hybrid reactor (FDS-SFB), has been chosen as the computational model. All the designs and calculations have been carried out by using SuperMC Monte-Carlo code with ENDF/B-VII neutron cross-section library. Beryllium (Be), beryllium oxide (BeO), HT-9 stainless steel, tungsten carbide (WC), magnesium oxide (MgO), silicon carbide (SiC), and titanium carbide (TiC) have been selected as the candidate reflectors. Other neutronic parameters, such as the neutron flux spectrum as well as the effective multiplication factor (keff), have been calculated for each of the reflectors to understand the behavior of each reflector material. In the case of tritium production, as expected, Be and BeO had the best performance and if a non-beryllium reflector was considered, MgO could be the best. In the case of transmutation of TRUs, for 241Am, 243Am, and 242Pu, BeO had the best results. For 244Cm and 237Np, TiC and MgO were the best choices respectively. For the total transmutation, BeO, MgO, and WC had the best capability to transmute TRUs among the other reflectors. Considering both parameters, i.e., TRUs transmutation and Tritium production, BeO (among beryllium-reflectors) and MgO (among non-beryllium reflectors) could be an appropriate choice to fusion hybrid designs.

The current manuscript presents a conceptual analysis of the hybrid tokamak operating with D–T plasmas in regimes with high values of the fusion gain Q (Q = fusion power output/auxiliary heating power input) or even in the ignition regime. The total fusion power release can reach $$P_{\text {fus}}$$ = 150–200 MW. The key features of the hybrid system fusion–fission are a relatively small size and a construction simplicity of this plasma device (minor and major radii of the tokamak are $$a = 1.2$$ m and $$R = 4.4$$ m, respectively). It is assumed that the hybrid tokamak will operate in the so-called radiative improved (RI-)mode which has been extensively investigated in the tokamak TEXTOR-94 and other machines. This regime allows to overcome a range of practical difficulties existing nowadays to tokamak-reactors (for instance, the power exhaust). Additionally, the RI-mode gives a possibility to extend the operational limits, in particular, to increase considerably the plasma density. The main parameters of the designed tokamak have been obtained using a simple system code. The critical issues of using such a tokamak as a basis for a fusion neutron source or for a hybrid reactor device are discussed.

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H2O, D2O, Na, PbBi, LiF-BeF2 and NaF-BeF2 applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are keff, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H2O and D2O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF2 and NaF-BeF2 is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

In the present investigation, a hybrid reactor system was designed by using 100 % Flibe, 90 % Flibe-10 % ThF4, 90 % Flibe-10 % UF4 fluids, ENDF/B-VII, JEFF-3.1, JENDL-4.0, ROSFOND, BROND-2.2, CENDL-3.1 evaluated nuclear data libraries and Ferritic Steel, 9Cr2WVTa, V4Cr4Ti, SiC structural materials. The fluids were used in the liquid first wall, liquid second wall (blanket) and shield zones of a fusion–fission hybrid reactor system. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional nucleonic calculations were performed using the most recent version MCNPX-2.7.0 the Monte Carlo code.

A team at the Budker Institute of Nuclear Physics has been working for several years to develop the Gas Dynamic Trap Mirror Neutron Source (GDT-NS) for fusion materials irradiation. In 2010 they op...

Fusion-fission hybrid reactor (FFHR), a representative kind of subcritical nuclear reactor, has the advantage of better safety margin. There are less constraints for the loading of active fuel when burning spent nuclear fuel. A simplified three-dimensional model of tokamak blanket is constructed by the Serpent Monte Carlo Method Code to investigate the MA transmutation characteristics in the Flibe molten salt blanket. Based on the model with one homogenous fuel region, the effect of key design parameters: MA content, plutonium content, Li-6 enrichment in the initial fuel were determined, during which performance characteristics such as MA transmutation amount, keff, tritium breeding ratio (TBR), blanket energy multiplication factor M and conversion ratio were obtained. The results show that it is hard to improve the tritium breeding and meanwhile enhance the MA reducing in one homogenous fuel region. To balance these two functions, a double molten salt fuel regions model with separative MA transmutation zone and tritium breeding zone was proposed.

World is facing issues regarding waste management of fission reactors and this idea of fusion-fission hybrid reactor is very attractive solution of waste burning with evolution of fusion energy. In this paper, concept of clean nuclear power plant is proposed which is done by burning all high level waste, it include transuranic (TRU) isotopes in a fusion-fission hybrid reactor and FP region. Fusion based TRU burner reactor (FTBR) is proposed which contains fixed power tokamak and ring shaped annular TRU core with metal fuel. The power level of TRU core is set to be 2000 MW. Small tokamak geometry with a plate type core is designed to observe TRU and FP transmutation. Sub-criticality calculations were done. Simulations were performed using SuperMC and flux was calculated as a function of energy and spatial variation. Long lived radionuclides were transmuted into short lived ones.

Attaining tritium self-sufficiency is indispensable in a Z-pinch-driven fusion–fission hybrid reactor (Z-FFR). In this paper, a conceptual design is presented in which the Z-FFR tritium cycle system was divided into eight subsystems. A theoretical analysis of tritium inventory based on the mean residence time was performed to quantitatively obtain the tritium distribution in each subsystem. Tritium self-sufficiency judgment criteria were established using a tritium mass flow analysis method. The dependency relationships between the burning rate, tritium breeding ratio, extraction efficiency, and tritium self-sufficiency were also specified for the steady state.

Outcomes from RFX-mod, the largest Reversed Field Pinch (RFP) experiment, have shown a substantial increase of the electron and ion temperatures versus the plasma current with a trend that allows hypothesizing significant D-T fusion processes in larger devices with higher plasma current. Such a device, in which the plasma is purely ohmically heated, could act as an efficient, robust and cheap fusion neutron source with a neutron rate to be used as the basis for a Fusion-Fission Hybrid Reactor (FFHR). The peculiar features of this neutron source can be summarized as: ohmically heated plasma, toroidal field winding rated for a low magnetic field (one-two orders of magnitude lower than in Tokamaks), very simple, robust and cheap construction, simple access for Remote Handling and easy maintenance. Almost continuous neutron production, as required in a hybrid reactor, is guaranteed by a continuously pulsed operation without the need of additional current drive systems. The aim of the paper is to investigate the relationship between the machine size (major and minor device radius), the attainable stored volt-seconds and, consequently, the maximum plasma current and pulse duration. The analysis uses the experimental RFX-mod data (electron temperatures, plasma currents, loop voltages during flat top) to extrapolate the performances of larger RFP devices with significant neutron production from D-T reactions. This study shows a realistic possibility, in an inductively operated RFP of larger size than RFX-mod, with plasma current 15-20 MA and temperature approximate to 10-15 keV, to reach the level of neutron generation of a fusion fission hybrid reactor, leading to a neutron production rate in the range of 10(19) n/s and a wall neutron load of 0.2 MW/m(2)

Summary One of the most important parameters in the design of the fusion-fission hybrid reactor is the selection of the first wall material. Because the oxide dispersion-strengthened (ODS) steel alloys have high temperature oxidation, high radiation resistance, good hardness, and corrosion resistance properties, they are thought first wall candidate materials for fusion and fission applications. The objective of this paper is to determine the best radiation damage parameters of various experimental and commercial ODS steels (namely, 12Y1, 12YWT, 1DS, IDK, Eurofer97, MA956, MA957, and PM 2000). Neutron spectrum and average neutron energy throughout blanket, displacement per atom, hydrogen and helium production, nuclear heating, and tritium breeding ratio were calculated by using Monte Carlo methods with Monte Carlo neutron-photon transport code and nuclear libraries named as ENDF/B-VI and CLAW-IV. It is assumed that calculated reactor has been operated full power during a year and neutron wall load is 2.25 MW/m2 (1014 n/s). All investigated first wall materials should be replaced between 3.5 and 4 years. All investigated materials provide minimum required tritium breeding ratio value, and when considering all the calculations performed in this work, 1DS ODS steel is the most suitable first wall materials with respect to other investigated ODS steels.

In this study, salt-heavy metal mixtures consisting of 93–85% Li20Sn80 + 5% SFG-PuO2 and 2–10% UO2, 93–85% Li20Sn80 + 5% SFG-PuO2 and 2–10% NpO2, and 93–85% Li20Sn80 + 5% SFG-PuO2 and 2–10% UCO were used as fluids. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion–fission hybrid reactor system. A beryllium (Be) zone with a width of 3 cm was used for neutron multiplicity between the liquid first wall and the blanket. 9Cr2WVTa ferritic steel with the width of 4 cm was used as the structural material. The contributions of each isotope in the fluids to the nuclear parameters, such as tritium breeding ratio (TBR), energy multiplication factor (M), and heat deposition rate, of the fusion–fission hybrid reactor were calculated in the liquid first wall, blanket, and shield zones. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

The paper describes recent progress in integral neutronics experiments in the analytical mockups for the blanket in a fusion-fission hybrid energy reactor. A conceptual blanket of the hybrid reactor is mainly loaded with natural uranium and lithium material. In the fission fuel region, uranium material and light water are arranged alternately. The mockups of the conceptual blanket are designed and used for checking neutron property of the blanket by integral experiments. Based on materials available, the spherical fission mockup for fission research and plutonium production consists of three layers of depleted uranium shells and several layers of polyethylene and graphite shells. The spherical lithium mockup for tritium production consists of depleted uranium and LiPb alloy shells. The cubic mockup consists of natural uranium and polyethylene and its structure is basically consistent with one of the fuel region. In the mockups with the D-T neutron source, the plutonium production rates, uranium fission rates and tritium production rates are measured, separately. The measured results are compared to the calculated ones with MCNP-4B code and ENDF/B-VI library data.

In this paper a hybrid chemical reactor, known also as the partially simulated reactor, is presented. Such reactor consists of a jacketed vessel and LabVIEW based real-time simulator of a chemical reaction. The reactor can be used for testing advanced control strategies. Experimenting with the hybrid reactor excludes risk of a thermal runaway because the simulated chemical reaction can be stopped at any time. To make the hybrid reactor a highly nonlinear benchmark plant, the reaction parameters must be properly selected, taking into account all the constraints of the designed system. For this reason, preliminary tests using a numerical simulator were performed to avoid time consuming experiments with the hybrid reactor. In our research, we consider three different models of the batch hybrid reactor. Results show that an overall heat transfer coefficient cannot be assumed constant to obtain a good agreement between the simulation and the measurement data. Moreover, the overall heat transfer coefficient varies with flow through the jacket and a dispersion of its values is different for each model.

The reactor studied in this work is the hybrid fusion–fission transmutation system (FFTS), which is a fusion–fission hybrid reactor with a central compact fusion neutron source (CFNS). It is based on the Tokamak concept, and it is surrounded by a zone made of transuranic elements obtained from reprocessing and recycling of spent fuel of light water reactors. High-energy neutrons, of fourteen MeV, are generated in the CFNS; they are produced by the deuterium-tritium reaction. In this study, the MCNPX Monte Carlo code was used to build up a model of the FFTS for studying the tritium breeding capability of the system. Tritium is produced from neutron capture in lithium, which is located in blankets specifically designed for this purpose. The tritium breeding ratio (TBR) is defined as the average number of tritium atoms bred per tritium atom burnt in the deuterium-tritium reaction. We must have TBR>1, for a self-sustained fusion economy. In the first step of this work, the location of the lithium blankets was defined. Afterwards, different blanket materials were tested: natural lithium, enriched lithium in 6Li, different lithium alloys with neutron multipliers like lead and beryllium [Li4SiO4, LiTiO3, FLiNaBe, FLiBe, Pb-15.8Li (Li-6 at 90 %)]. Finally, a study was carried out to determine the relationship between the width of the blanket and the tritium breeding. Concerning the blanket locations, we defined four: one in the central column of the FFTS, one in the upper and one in the bottom part of the fusion region of the system, and the last one in the external part of the fission region. This means that the first three blankets use high-energy neutrons from the deuterium-tritium reaction, and the fourth blanket uses neutron leaking from the fission reactions. The principal results show that the best option is the blanket with Pb-15.8Li with lithium enriched at 90 % in lithium-6 with TBR = 1.09. It was found that the blanket at the external part of the fission region has the higher tritium breeding capability. Regarding the blanket width, it was observed that most of the tritium breeding is carried out in the first 5 cm of the blanket, and beyond this width breeding is minimal; therefore, for the blanket it is more important to have a high view factor to neutrons (i.e., a big surface exposed to neutrons) than a deep region. Finally, it is important to mention that the FFTS was critical during 1000 days that were simulated with MCNPX.

In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99–95% Li20 Sn80 + 1-5% RG-Pu, 99–95% Li20 Sn80 + 1-5% RG-PuF4 , and 99–95% Li20 Sn80 + 1-5% RG-PuO2 , as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion–fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR ), energy multiplication factor (M ), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency ( TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

In this study, the molten salt-heavy metal mixtures 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % UO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % NpO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion–fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.

In the present investigation, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa ferritic steel structural material and 99-95 % Li20Sn80-1-5 % SFG-Pu, 99-95 % Li20Sn80-1-5 % SFG-PuF4, 99-95 % Li20Sn80-1-5 % SFG-PuO2 the molten salt-heavy metal mixtures, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium zone with the width of 3 cm was used for the neutron multiplicity between liquid first wall and blanket. The contributions of each isotope in fluids on the nuclear parameters of a fusion-fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, heat deposition rate were computed in liquid first wall, blanket and shield zones. Three-dimensional analyses were performed by using Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs.

This study proposes a conceptual design of a hybrid reactor for waste transmutation (Hyb-WT). The design of Hyb-WT is based on a low-power tokamak (less than 150 MWt) and an annular ring-shaped reactor core with metal fuel (TRU 60 w/o, Zr 40 w/o) and a fission product (FP) zone. The computational code systems MONTEBURNS and MCNPX2.6 are investigated for their suitability in evaluating the performance of Hyb-WT. The overall design performance of the proposed reactor is determined by considering pin-type and tube-in-duct core designs. The objective of such consideration is to explore the possibilities for enhanced transmutation with reduced wall loading from fusion neutrons and reduced transuranic (TRU) inventory. TRU and FP depletion is analyzed by calculating waste transmutation ratio, mass burned per full power year (in units of kg/fpy), and support ratio. The radio toxicity analysis of TRUs and FPs is performed by calculating the percentage of toxicity reduction in TRU and FP over a burn cycle.

*The blanket materials were evaluated and optimized for use in a nuclear fusion-fission hybrid reactor by calculating the tritium breeding ratio (TBR), the ratio of thermal power output to fusion power input (Q value) and the temperature distributions. The calculation of TBR and Q value was performed by Monte Carlo code (MCNP-4C) and the temperature distributions by one dimensional heat equation. First, the structure materials, coolant and tritium breeding materials were investigated in terms of the TBR and the Q value. Secondly, the layer structure was optimized by calculating the temperature distributions in the blanket where the materials were selected by considering the former results of the TBR and Q value. As a result, the highest TBR and Q value could be obtained, if vanadium alloy structure material, helium gas coolant and molten salt tritium breeding material would be used as the blanket material. In addition, by optimizing the blanket structure using these materials, it would be possible to get a high TBR of 2.0 maintaining the temperature limit of the blanket materials, and about twenty times the thermal power output could be produced compared to the fusion power input under the condition of effective multiplication factor of 0.80.