Fusion-fission Hybrid Energy Reactor Research Articles

Simulation of loss-of-flow transient in water cooled fission blanket of Fusion-fission Hybrid Energy Reactor (FFHER)

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.

Neutronic Study of an Innovative Thorium-Uranium–Based Fusion-Fission Hybrid Energy Reactor with 233U Breeding Enhancement by Using Dual-Coolant System

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 thorium module. The uranium and thorium modules have weak coupling in neutronic behavior. However, with the fixed design parameters of the uranium and thorium modules, the most important influencing factor on energy multiplication factor M (the ratio of total blanket energy output and the fusion energy) and the 233U breeding rate is the fraction of the external fusion neutron source irradiated on the uranium or thorium module or the blanket coverage rate of the uranium or thorium modules. Based on this characteristic, an innovative hybrid reactor design that employs a dual-coolant system is proposed in this technical note. Uranium modules still use water as the coolant to maintain a high energy multiplication factor, whereas helium is used as the coolant for the thorium module to obtain a fast neutron spectrum to enhance the 233U breeding. The simulation results show that the helium-cooled thorium module is 2.5 times more efficient in 233U breeding compared to the original water-cooled thorium module design. Approximately 10 tons of 233U is produced after 20 years of operation for the helium-cooled thorium module design.

Shielding Design and Analysis of Fusion–Fission Hybrid Energy Reactor Blanket

A preliminary design of fusion–fission hybrid energy reactor (FFHER) has been proposed by Institute of Nuclear Physics and Chemistry based on current fusion science and well-developed fission technology. In FFHER, shield blocks provide nuclear shielding and thermal shielding for internal and external blanket components. The hybrid of fusion core and fission blanket makes the spectra rather complex. Therefore, it is necessary to make detail shielding design and carry out radiation analysis according to the blanket structure and material property. In this study, a shielding design of combining several different material shield blocks has been proposed. The shielding analysis is performed by Monte Carlo (MC) method. For the radiation deep-penetration problem, the flux and statistical relative error of forward MC estimate are applied to get an optimal weight window for global variance reduction (GVR). The spatial distribution of neutron and gamma flux have been assessed along the shield block depth. Power deposited and dose rate distributions have also been simulated and analysed. Neutron irradiation damage has been studied to evaluate the material damage. Based on the configuration analysis, nuclear analysis and GVR method, an optimal FFHER blanket shielding design has been obtained.

Experimental and simulative studies of thorium fission rates on thorium oxide/depleted uranium cylinders with D-T neutrons

Aiming at validating thorium fission parameters and the code used in the conceptual design of fusion-fission hybrid energy reactor, an integral experiment on thorium oxide/depleted uranium cylinder assemblies for thorium fission rate has been carried out with D-T neutrons. Thorium fission rate in the axial direction of the assemblies was obtained by adopting the activation method, and the experimental uncertainties were 5.6–5.9%. MCNP simulation employing ENDF/B-VII.0 and ENDF/B-VII.1 library data were 5%–21% smaller than experimental ones. The influence of the polyethylene oxide (PEO) in the thorium oxide cylinders, evaluated by MCNP simulation employing ENDF/B-VII.0 library data, is negligible in the present thorium/uranium assemblies under the measured level. The difference between the present results and our previous ones was also analyzed. The stage experimental results could provide reference for the evaluation of neutron-induced thorium fission cross section, and should be considered in the margin of the conceptual design of FFHER.

Measurement of 232Th(n,f) and 232Th(n,γ) reaction rates in thorium powder cylinder bombarded with D-T neutrons

ABSTRACTIn order to provide reference for the evaluation of thorium parameters for the conceptual design of fusion–fission hybrid energy reactor, a dedicated integral experiment was carried out in a thorium powder cylinder bombarded with D-T neutrons. Thorium capture and fission rates in the 0° direction to the incident D+ beam were obtained using the activation method followed by the off-line gamma-ray technique, experimental uncertainties were ∼3.1% for thorium capture rate, and were 5.5%, 8.1%, and 6.3% for thorium fission rates based on fission products 85mKr, 143Ce, and 87Kr, respectively. The thorium fission rate based on 85mKr agreed well with the simulation employing ENDF/B-VII.0 library data. The influence of the oxygen contained in the thorium oxide powder and the scattering neutrons from the experimental hall was also evaluated. MCNP simulations employing ENDF/B-Ⅶ.0, JENDL-4.0 library data agreed with experiment within uncertainties except that employing ENDF/B-Ⅶ.1 (6.0%) and CENDL-3.1 (7.9%) for thorium fission rate, while for thorium capture rate, simulation employing JENDL-4.0 agreed with experiment best. The influence of reaction channels of thorium transport medium employing different library data on the thorium reaction rates could be neglected according to the simulation. The thorium capture to fission ratio demonstrated that the fuel breeding efficiency is quite low and energy production plays a leading role under the neutron spectra in this experiment.

Measurement and calculation of U and Th reaction rates in uranium mock assemblies

The conceptual blanket based on a subcritical reactor for a fusion–fission hybrid energy reactor with magnetic confinement fusion is being studied. To validate nuclear data and code of neutronics calculation for the blanket design, measurement of reaction rates for U and Th is performed by integral experiments. Based on the fission blanket, the uranium mock assemblies are established, and include the depleted uranium shell, the depleted-uranium/polyethylene/graphite shell and the natural uranium cube. A depleted-uranium/polyethylene/graphite assembly of 131mm in inner radius and 365mm in outer radius consists of three layers of depleted uranium shells, two layers of polyethylene shells and a graphite shell, and these shells are arranged alternatively. A D–T fusion neutron source is centered of the spherical assemblies. The U reaction rates for capture, fission and ratio of capture to fission in the assemblies are measured by depleted uranium foils and an HPGe gamma spectrometer. The 232Th reaction rates for capture, fission and ratio of capture to fission are measured by ThO2 foils put into the depleted uranium shell. The measured results are compared to the calculated ones by using MCNP code with ENDF/B-VII.0 library data.

Measurement and analysis of fission rates in depleted uranium and depleted uranium–polyethylene shells with neutrons from D-T reaction

To validate the concept design of a novel fusion–fission hybrid energy reactor, a depleted uranium assembly and a combined assembly of uranium and polyethylene were designed and assembled based on a depleted uranium spherical shell and a polyethylene spherical shell. The distribution of the fission rates for the depleted uranium and enriched uranium in the two assemblies, as a function of the distance of the detection position to the centre, was measured using a plate fission chamber bombarded by D-T neutrons. The addition of a polyethylene shell significantly changed the neutron spectrum; in particular, the neutron fluxes with energies of 1 MeV and lower were changed. Using MCNP5 and the attached libraries, the fission rate experiments were simulated, and the experimental configuration, including the wall of the experimental hall, was described in detail in the model. The fission rate distributions for depleted uranium and enriched uranium in the two assemblies were reproducible. The difference between the calculated results with different libraries and different tallies is as small as 1.0%. By considering the neutron flux, the fission rate and the C/E values, it is concluded that the fission rates of depleted uranium and enriched uranium induced by the fast neutrons were overestimated, and it is proposed that the fission parameters of uranium for fast neutrons should be re-evaluated, or the margin of the concept design should be enlarged, to make the concept effective.

Transient heat transfer performance of Z-pinch driven fusion-fission hybrid energy reactor blanket

Z-pinch driven fusion-fission hybrid energy reactor (Z-FFR) is a new energy system, which is helpful to the sustainable development of nuclear energy. It runs in a pulsed mode with a period of 10s, in order to output a mean thermal power of 3000 MWt, a large quantity of energy has to be generated in a pulse. The transient heat transfer and temperature variation of the blanket and first wall in the strong heat shock situation are key issues to the technical feasibility of Z-FFR. In this paper, the temperature variation of blanket and first wall under consecutive heat pulse is analyzed through numerical simulation. Meanwhile, with the aim of exporting electric power stably, a simple method of stabilizing the output power is proposed, and the fluctuation character of the outflow in the main pipe is described. The results show that, the maximum temperatures of all materials are in the safety margin, the thickness of the layer that has a relative high temperature on the surface of the first wall is about 0.5 mm, the output power has a fluctuation of -2.84% to +2.05% relative to the nominal value.

Z箍缩驱动混合堆包层瞬态传热特性

Z箍缩驱动混合堆包层瞬态传热特性

Progress in physics design of fusion-fission hybrid energy reactor

In this paper, we propose a preliminary design for a fusion-fission hybrid energy reactor (FFHER), based on current fusion science and technology and well-developed fission technology. Design rules are listed and a primary concept blanket with uranium alloy as fuel and water as coolant is put forward. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. The interaction between the fusion neutron and the uranium fuel with the aim of achieving greater energy multiplication and tritium sustainability is studied. Other concept hybrid reactor designs are also reviewed. Integral neutron experiments were carried out to verify the credibility of our proposed physical design. The combination of the physical design with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology basis for future development of the FFHER concept in China.

Integral neutronics experiments in analytical mockups for blanket of a hybrid reactor

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.

次临界能源堆物理设计进展

In this paper, we propose a preliminary design for a fusion-fission hybrid energy reactor (FFHER), based on current fusion science and technology and well-developed fission technology. Design rules are listed and a primary concept blanket with uranium alloy as fuel and water as coolant is put forward. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. The interaction between the fusion neutron and the uranium fuel with the aim of achieving greater energy multiplication and tritium sustainability is studied. Other concept hybrid reactor designs are also reviewed. Integral neutron experiments were carried out to verify the credibility of our proposed physical design. The combination of the physical design with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology basis for future development of the FFHER concept in China.

Preliminary design of hybrid energy reactor and integral neutron experiments

We propose a preliminary design for a fusion–fission hybrid energy reactor (FFHER), based on current fusion science and technology (with some extrapolations forward from ITER) and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. We study the interaction between the fusion neutron and the uranium fuel with the aim of to achieving greater energy multiplication and tritium sustainability. We also review other concept hybrid reactor designs. We design integral neutron experiments in order to verify the credibility of our proposed physical design. The combination of this program of research with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology foundation for the future development of the FFHER concept in China.

Preliminary Design of Hybrid Energy Reactor and Related Nuclear Fuel Cycle

We propose a preliminary design for a fusion-fission hybrid energy reactor (FFHER), based on current fusion science and technology and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The FFHER could achieve greater energy multiplication (M>10 for U-Zr alloy fuel and M>5 for UO2 fuel) and tritium sustainability (TBR>1.05). The sub-critical blanket will last 30 years without reshuffling fuel. Fission products are the only waste that needs disposal. A new dry process called Fission Product Removal (FPR) replaces conventional reprocessing. It is only necessary to remove the cladding, vent the volatiles and pulverize the solids as feedstock for EM2 fuel fabrication. The AIROX (or DUPIC) process is an example of this operation and has been well demonstrated. After removing the fission products from its 30-year discharge, the refabricated fuel is returned to the reactor for another cycle, thereby reducing the need for enrichment and the proliferation resistance would be increased.

The Project of Fusion-Fission Hybrid Energy Reactor in China

The project, conceptual design of the fusion-fission hybrid energy reactors (FFHER) and related verifying experiments, was started in the framework of the National Magnetic Confinement Fusion Science Program of China in May 2010.Based on the current or slightly extrapolated ITER fusion science and technology and the well developed fission technology, fusion-driven sub-critical systems design was performed. We put forward the main concept of FFHER blanket with uranium alloy as fuel and water as coolant. The uranium can be natural uranium, LWR spent fuel or depleted uranium. FFHER can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy.We study the interaction among the fusion neutron and fuel to achieve larger energy multiplication and tritium sustainable, verify the creditability of physical design by the integral neutron experiments. With the combination research of this program and the following thermal hydraulic design, alloy fuel manufacture and nuclear fuel cycle programs provide the science and technology foundation for the future development of FFHER in China.