132 results about "Tritium" patented technology

Main characteristic of the invention is the exploitation of the very large flux of high energy neutrons generated in a nuclear fusion tokamak with beam-driven plasma for the purpose of producing radioisotopes. It specifically relates to the deployment of precursor materials into the core of a nuclear fusion tokamak with beam-driven plasma for the purpose of exposing these materials to the neutron flux generated in the plasma by the process of nuclear fusion of Deuterium (D) and Tritium (T) to generate radioactive isotopes, so-called radioisotopes. The precursor materials contained in metallic capsules are incorporated into the blankets that surround the plasma.

A mobile fusion device (MFD) comprises three main sections: the MFD reactor (467), MFD support sub-systems (468), and the MFD support and fuel injection sub-systems (469), which are indicated by broken lines in Figure 1 The MFD reactor (467) is where the Deuterium and Tritium fuel is heated to high enough temperatures and densities and confinement times that fusion occurs, releasing neutrons that are absorbed in and heats up the 17Li-83Pb breeder blanket where more Tritium fuel is bred and where heat is transferred to Helium coolant. The MFD support sub-systems (468) houses auxiliary power units that generate electricity to power the electrical sub-systems of the MFD, and houses the MFD control computer, and where the Helium coolant is cryogenically cooled and purified, and where the 17Li-83Pb breeder blanket fluid is purified by separating and storing Deuterium fuel and Tritium fuel and Helium coolant, and where a Helium heat exchanger is located. The MFD support and fuel injection sub-systems (469) houses Deuterium and Tritium neutral beam injection plasma heating systems and a Deuterium and Tritium fuel gas puffer system and a Deuterium and Tritium fuel storage and pellet manufacturing system, and where the 17Li-83Pb breeder blanket fluid is purified by separating and storing Deuterium fuel and Tritium fuel and Helium coolant, and where a Helium heat exchanger is located, and where the Helium coolant is cryogenically cooled and purified, and houses auxiliary power units that generate electricity to power the electrical sub-systems of the MFD. The MFD is housed within a multi-layered pressurised outer containment pod MLP, which has a multi-layered pressurised rear access hatch MLH. The MFD is mounted on a sliding raft SR for ease of access.

The invention discloses an online monitoring system for tritium in water based on a PSF array, and relates to the field of nuclear energy. The on-line monitoring system for tritium in water comprises a fluid control system, a PSF detector module and an electronics system, the fluid control system automatically controls to open electromagnetic valves of a sample inlet and a sample outlet, and a peristaltic pump is started, so that a water sample to be detected is pumped into a detection chamber of the PSF detector module through the sample inlet; after the detection chamber is filled with the to-be-detected water sample, the electromagnetic valve and the peristaltic pump are automatically controlled to be closed; the PSF detector module captures precipitation energy generated by tritium decay beta particles in a water sample to be detected through a PSF array, converts the precipitation energy into photon signals, and converts the photon signals into electric signals through a detection window and a double photomultiplier; and the electronics system performs signal preprocessing on the received electric signal, and obtains corresponding tritium content monitoring data to be detected based on the preprocessed electric signal. The on-line monitoring system for tritium in water can ensure that the finally obtained measurement result is more accurate.

The invention belongs to the technical field of tritium removal pump sets, and discloses a molecular pump set for evacuating a hydrogen isotope-containing super-large cavity, which comprises a stainless steel closed cavity and a super-large vacuum cavity, and is characterized in that a first tritium removal pump set and a second tritium removal pump set are respectively arranged in the stainless steel closed cavity through a fixing frame; the first tritium removal pump set and the second tritium removal pump set each comprise four tritium removal pumps, and the first tritium removal pump set and the second tritium removal pump set are integrally arranged in parallel and both communicate with the ultra-large vacuum cavity. According to the invention, the eight tritium removal pumps are mounted in the stainless steel chamber, and a connection mode of'parallel connection + series connection 'is creatively adopted, so that the requirement of continuous evacuation of the large chamber is met. In the initial stage, the eight tritium removal pumps operate in parallel so as to improve the overall pumping speed, the latter four tritium removal pumps serve as backing pumps, and the other four tritium removal pumps serve as main pumping pumps so as to improve the compression ratio of pumped tritium, increase the pressure of an exhaust port and improve the pumping efficiency of tritium.

The invention relates to the technical field of engineering logging, and discloses a low-power-consumption radio frequency ion source logging neutron tube and a working method, deuterium gas is ionized into deuterium ions through an ion source, the deuterium ions are accelerated under the action of a strong electric field to obtain energy, and the high-energy deuterium ions bombard a deuterium / tritium target to generate nuclear reaction to generate neutrons. The power requirement of the radio frequency source can be greatly reduced. Compared with an inductive coupling type radio frequency ion source neutron tube, the capacitive coupling type neutron tube has the advantages that D1 + 1 in plasma is increased to 50%, the required power is 25%-30% of that of the inductive coupling type neutron tube, the load of a neutron generator is greatly reduced, the D-D deuterium neutron tube is further applied to engineering practice, and powerful technical support is provided for green and environment-friendly well logging.

The application discloses a fusion reactor solid-state blanket design method based on topology optimization, and is characterized in that based on the topology structure characteristics of the fusion reactor solid-state blanket arranged in layers along the radial direction, the coupling mechanism of the multi-physical fields of neutronics, thermal-hydraulics and structural mechanics is researched, a coupling model between various parameters of the multi-physical fields is established through theoretical analysis and a multivariate nonlinear regression method; under the constraint conditions of material temperature and structural mechanical properties, a constrained topology optimization model for maximizing the tritium breeding ratio is constructed; the constrained topology optimization model is converted into an unconstrained topology optimization model for maximizing the tritium breeding ratio by using a Lagrange function and a KKT condition, the unconstrained topology optimization model is solved by using a gradient descent algorithm, and a blanket radial topology structure capable of meeting the design criteria is obtained. A blanket three-dimensional model is established, and the performance of neutronics, thermal-hydraulics and structural mechanics in the whole reactor is analyzed, the tritium production, heat removal and structural safety performance of the blanket are evaluated, and it is verified that the topology optimization model can efficiently obtain a blanket design scheme with optimal comprehensive performance of multiple fields.

This invention discloses a method for producing radioactive isotopes in the core of an advanced high-temperature reactor using a spherical target. This invention overcomes the limitations of using neutron irradiation channels or fuel-related components of nuclear reactors, such as conditioning rods in heavy water reactors and combustible poison rods in pressurized water reactors, to produce radioactive isotopes. It sets a precedent for the mass production of radioactive isotopes using the core of a nuclear reactor, and has significant strategic implications for exploring high-value applications of commercial advanced high-temperature reactors to reduce operating costs and mass-produce radioactive isotopes, including the strategic isotope tritium.

A controlled fusion process is provided that can produce a sustained series of fusion reactions: a process that (i) uses a substantially higher reactant density of the deuterium and tritium gases by converging cationic reactants into the higher reaction density at a target cathode rather than relying on random collisions, the converging producing a substantially higher rate of fusion and energy production; (ii) uses a substantially lower input of energy to initiate the fusion; (iii) can be cycled at a substantially higher cycle frequency; (iv) has a practical heat exchange method; (v) is substantially less costly to manufacture, operate, and maintain; and, (vi) has a substantially improved reaction efficiency as a result of not mixing reactants with products.

The application provides an automatic sampling and measuring device for radioactive aerosol and gaseous tritium in radioactive gaseous effluent. The measuring device comprises a radioactive aerosol measuring module and a gaseous tritium measuring module integrated in the measuring device, and the radioactive aerosol measuring module and the gaseous tritium measuring module are communicated with a gaseous effluent pipeline through sampling pipelines. The measuring device of the application can greatly save labor, realize unattended and long-time online monitoring, flexibly adjust the sampling and monitoring frequency, and improve the monitoring efficiency. In addition, in the case of a nuclear accident, the emergency monitoring mode can be quickly started to quickly respond to the accident monitoring requirements, so as to realize continuous, online and automatic monitoring of the concentration of radioactive aerosol and gaseous tritium in the effluent of a nuclear facility.

The present invention discloses a system and method for measuring the deuterium-tritium fuel ratio in plasma, comprising an electron cyclotron radiometer, a microwave reflectometer, a Mirnov probe, electron cyclotron radiation imaging, a first collector, a central processing unit, an interferometer, a dynamic Stark effect spectrometer, a magnetic coil, a charge exchange spectrometer, a second collector, a high-performance computer, and a TAE frequency real-time processor. The deuterium-tritium fuel ratio can be obtained by accurately measuring the frequency of the toroidal Alfvén eigenmode in the burning plasma and accurately calculating the TAE frequency in the deuterium-deuterium plasma under the same conditions. Compared to the existing technology, even if the burning plasma is in an uneven distribution state, it can be measured as a whole, and the deuterium-tritium fuel ratio can be calculated based on the measured data, thereby providing a safe, accurate, and rapid method for measuring the deuterium-tritium fuel ratio for the heating effect of fusion product alpha particles, the high and low confinement mode conversion power thresholds, and the confinement time.

The present invention relates to the field of nuclear radiation detection technology, and provides a tritium measurement device with radon discrimination capabilities. The device comprises a gas flow ionization chamber, a current amplification circuit, a pulse amplification circuit, and a signal processing circuit. The collector of the gas flow ionization chamber is connected to the current amplification circuit, which amplifies and measures the current signal in the gas flow ionization chamber. The measured signal includes the current signals of tritium beta particles and radon alpha particles. The high-voltage electrode of the gas flow ionization chamber is connected to the pulse amplification circuit after passing through a DC-blocking capacitor, which measures the pulse signal generated by radon alpha particles. The device structure of the present invention is convenient to implement, comprising only a single detection device. It can measure and discriminate radon alpha particles entering the gas flow tritium ionization chamber in real time, accurately compensate for tritium measurement results, and prevent further increase in the total mean square error of the measurement results after compensation.

The invention discloses a neutron-gamma mixed radiation screening method based on a radiation response signal. The neutron-gamma mixed radiation screening method depends on a nano-coating CMOS sensor neutron measurement system. The neutron measurement system of the nano-coating CMOS sensor comprises a detector, the detector comprises a CMOS sensor; a LiF thin film is attached to the light-sensitive surface of the CMOS sensor; the method comprises the following steps that a part of neutrons enable the CMOS sensor to generate a radiation response signal A; the other part of neutrons and 6Li are subjected to nuclear reaction to generate alpha rays and tritium ions, so that the CMOS sensor respectively generates a radiation response signal B and a radiation response signal C; the gamma ray enables the CMOS sensor to generate a radiation response signal D; and according to the difference of the four radiation response signals in morphology and / or gray values, discriminating the neutron-gamma mixed radiation field. The detector formed by combining the LiF film and the CMOS sensor is used for discriminating a neutron-gamma mixed radiation field and can be used for monitoring nuclear leakage of a reactor.

The invention discloses an alloy for hydrogen isotope gas storage and a preparation method thereof. The chemical formula of the hydrogen isotope gas storage alloy is ZrMn2-x, wherein x is more than or equal to 0.25 and less than or equal to 0.5. Compared with the prior art, the alloy has the following beneficial effects: 1, the alloy prepared by the invention has lower hydrogen absorption / desorption isotope equilibrium pressure, and when the alloy is used for tritium gas storage, the quantity of tritium remaining in a gas phase in the storage process can be greatly reduced, so that helium generated by decay of tritium remaining in the gas phase is reduced to the greatest extent; and 2, the prepared alloy does not contain catalytic active elements, and generation of tritiated methane gas in the tritium release process of the tritium storage alloy is avoided, so that the chemical purity of tritium gas is improved. And 3, the main element of the alloy is the low-cost Mn element, so that the cost of the hydrogen isotope gas storage alloy can be greatly reduced.

The invention discloses a method for producing radioactive isotopes in a reactor core of an advanced high-temperature reactor by adopting a spherical target piece. According to the invention, the limitation that the radioactive isotope can only be produced by using a neutron irradiation channel of a nuclear reactor or fuel-related components such as an adjusting rod of a heavy water reactor and a burnable poison rod of a pressurized water reactor is broken through, and a precedent of mass production of the radioactive isotope by using the reactor core of the nuclear reactor is created; the method has important strategic significance for exploring high-value application scenes of commercial advanced high-temperature reactors to reduce operation cost and producing radioactive isotopes including strategic isotope tritium in batches.

This invention relates to a vertical oxidation-combustion collection system for sediment or biological radioactive organic carbon-tritium samples and its method of use, comprising: a vertical oxidation-combustion device for oxidizing and combusting the sample and catalyzing it into products to be separated; and a sample collection device comprising: a water vapor trapping module for collecting tritium samples; and a carbon trapping module for collecting... 14 Sample C; A rear-mounted protection module is used to discharge exhaust gas, collect waste liquid, and prevent external air interference with sample collection. This invention, by placing the collection port below the collection container, allows for the collection of tritium samples from combustion products under gravity. 14 Sample C is collected more thoroughly and can also prevent tritium from entering. 14 C-sample collection module, to avoid 14 Sample C enters the post-protection module; more importantly, nitrogen blowing or inert gas blowing can completely remove and collect any residual tritium sample in the collection module. 14 Like C, avoid the occurrence of the memory effect.

A metal-organic framework coating may be employed in a molten salt reactor system to capture gaseous fission products produced as a consequence of operation. The metal-organic framework coating may include one or more metal-organic frameworks configured to capture certain fission products based on the need, for example tritium fluoride, tritium gas, and hydrogen fluoride. The one or more metal-organic frameworks may be immobilized onto a substrate and formed into a coating covering the substrate. In many cases, the substrate is the stainless-steel surfaces of reactor enclosures and vessels. In other cases, the substrate may be suspended into a headspace of a vessel of a salt-bearing component, or may be included in an off-gas processing system of a reactor system.

The application discloses a fusion reactor liquid blanket tritium breeding rate online regulation system simulation method and device, wherein the simulation device comprises a fusion blanket neutron transport module, a blanket breeding agent Monte Carlo-activation coupling burnup simulation module, a blanket coolant feeding module, a blanket tritium extraction module, a blanket coolant purification treatment module, a blanket tritium breeding rate calculation module and a blanket tritium breeding agent circulation system; the simulation device realizes the circulation of the blanket coolant nuclide based on the mass conservation equation and the burnup equation. The application is based on the Boltzmann neutron transport equation, the neutron flux of the blanket coolant is simulated by a Monte Carlo method, the time and space variation law of the chemical composition of the system material is obtained by solving the burnup equation, the replenishment demand of the blanket coolant system is calculated, and the influence of the replenishment composition on the blanket tritium breeding rate is simulated, so that the application can be widely applied to the fusion reactor liquid lithium lead blanket.

The invention relates to the technical field of isotope batteries and discloses a nanopore-containing power generation assembly. The nanopore-containing power generation assembly comprises a transduction layer, the loading layer is arranged on the surface of the transduction layer in advance, and the loading layer is provided with preset nanopores. The technical problem that in the preparation process of the radioactive source in the isotope battery, due to the fact that the hydrogen absorption temperature of a hydrogen storage material is high, large thermal mismatching exists between a transducer device and the hydrogen storage material, growth stress is generated in the phase change process, and consequently the transducer device may have serious structural damage can be solved; the loading layer arranged on the surface of the transduction layer in advance is adopted, and the loading layer is provided with the nanopore structure, so that when the loading layer serves as a basic structure for preparing the isotope power generation assembly, for example, nanopores are filled with radioactive gas such as tritium, the loading layer structure is stable, and the isotope power generation assembly is prepared. The probability that the energy conversion layer structure is damaged due to growth stress generated in the phase change process can be reduced, and then it is guaranteed that the structure of the prepared isotope power generation assembly is stable.

The invention discloses a thermonuclear fusion reaction combustion product trapping method and system, and belongs to the field of nuclear fusion. The invention discloses a thermonuclear fusion reaction combustion product trapping method. The method comprises the following steps: S1, cooling a combustion product; s2, carrying out primary photolysis treatment on the cooled combustion product; s3, carrying out primary separation treatment on the combustion product subjected to the primary photolysis treatment; s4, carrying out first collection treatment on a first product obtained by the first separation treatment; s5, performing secondary photolysis treatment on the residual combustion product after the primary separation treatment; s6, carrying out second-time separation treatment on the residual combustion product after the second-time photolysis treatment; s7, collecting a second product obtained by the second separation treatment; and S8, residual combustion products obtained after the second-time separation treatment are collected and treated. According to the invention, mixed products with insufficient combustion of deuterium and tritium in the thermonuclear fusion reaction can be efficiently classified and collected.

The invention provides an automatic sampling and measuring method for gaseous tritium, which comprises the following steps: S1, sampling of tritium-containing gas: dissolving the tritium-containing gas in sampling liquid through a tritium-containing gas sampling unit to prepare tritium-containing liquid; s2, tritium-containing liquid sample preparation and sample feeding: tritium-containing liquid and scintillation liquid are proportionally injected into a measuring bottle through a tritium-containing liquid sample preparation unit, and then the moving position of the measuring bottle is controlled through a sample feeding operation unit; and S3, automatic measurement: measuring the flicker spectrum of the measurement bottle placed in an automatic measurement unit through the automatic measurement unit, and calculating the activity of tritium after measurement. According to the invention, automatic sampling, automatic sample preparation, automatic sample feeding and automatic measurement of tritium in air can be realized, automatic online monitoring of gaseous tritium is realized, and improvement of monitoring efficiency is facilitated; unattended operation and long-time full-automatic sampling and measurement of tritium in the atmospheric environment and gaseous effluent can be realized, and on-line monitoring of low-concentration tritium in air is realized; the system can be quickly started when a nuclear radiation accident occurs, and the response speed of nuclear emergency monitoring is improved.

This invention discloses a nano-alloy neutron target film material and its preparation method. The chemical formula of the target film material is Ti. a X b V c Al d Nb e Where X is selected from Zr or Mg, and a, b, c, d, and e represent the atomic percentages of the corresponding elements, with 5 ≤ a ≤ 40, 5 ≤ b ≤ 40, 5 ≤ c ≤ 40, 5 ≤ d ≤ 40, and 5 ≤ e ≤ 40, and a + b + c + d + e = 100. This invention, through rational composition design, coordination of multiple elements, and specific preparation methods, nanoscales high-entropy alloys, achieving a deuterium (tritium) release temperature of the neutron target exceeding 200°C, while simultaneously increasing the deuterium storage capacity H / M to over 1.8, effectively improving the temperature resistance and deuterium storage performance of the neutron target.

The invention belongs to the field of neutron tubes, and particularly relates to an alpha particle accompanying neutron tube for security check and a region measurement method based on the neutron tube. The neutron tube comprises a tube shell assembly, an ion source assembly, an acceleration collimation assembly, a deuterium tritium target assembly, a storage and an alpha particle detector, the ion source assembly is mounted at the upper end in the tube shell assembly; the acceleration collimation assembly is installed in the tube shell assembly and located on an output path of the ion source assembly. The deuterium tritium target assembly is installed at the lower end in the tube shell assembly and located on an output path of the acceleration collimation assembly. The storage device is installed at the lower end in the tube shell assembly. A mounting opening is formed in the lower end of the side wall of the tube shell assembly, the alpha particle detector is mounted at the mounting opening of the tube shell assembly, and the receiving surface of the alpha particle detector faces the target surface of the deuterium-tritium target body. According to the invention, the flight direction of neutrons can be determined by detecting the accompanying alpha particles, so that the purpose of regional detection is achieved.

The invention provides an automatic sampling, sample preparation and measurement device and method for radioactive aerosol and gaseous tritium in radioactive gaseous effluent. The measuring device comprises a radioactive aerosol measuring module and a gaseous tritium measuring module which are integrated in the measuring device, and the radioactive aerosol measuring module and the gaseous tritium measuring module are respectively communicated with a gaseous effluent pipeline. According to the measuring device, manpower is greatly saved, unattended and long-time online monitoring can be achieved, the sampling monitoring frequency can be flexibly adjusted, and the monitoring efficiency is improved. Besides, under the condition of a nuclear accident, an emergency monitoring mode can be quickly started, and the accident monitoring requirement can be quickly responded, so that continuous, online and automatic monitoring of the concentration of radioactive aerosol and gaseous tritium in the effluent of the nuclear facility is realized.

The present application particularly relates to a method for producing helium-3 from heavy water in a heavy water reactor, comprising: step 1, subjecting the heavy water in a moderator system of a heavy water reactor unit to negative pressure rectification to obtain concentrated high-tritium heavy water; step 2, subjecting the concentrated high-tritium heavy water in step 1 to beta decay to produce helium-3 raw gas; and step 3, subjecting the helium-3 raw gas in step 2 to active carbon adsorption, helium-argon separation and hydrogen isotope gas adsorption to obtain helium-3. The method for producing helium-3 from heavy water in a heavy water reactor avoids technical difficulties and high costs of helium-3 / helium-4 isotope separation, and has high technical maturity and good economy.

This invention discloses a method for accurately tracing geological hydrogen migration, belonging to the field of geological hydrogen exploration and migration tracing technology. The invention involves preparing stable hydrogen isotopes and radioactive tritium tracers with specific geochemical fingerprints and injecting them into target geological units; deploying a three-dimensional monitoring network including sensor arrays and automatic samplers to collect hydrogen concentration data and gas samples in real time; laboratory analysis to determine hydrogen isotope composition and rare gas concentrations, using an N2-He-Ar ternary diagram to identify gas sources and eliminate interference; integrating data and performing numerical simulations using migration control equations coupled with isotope fractionation effects, optimizing parameters through inversion algorithms; and generating a three-dimensional visualization map of hydrogen migration paths and a quantitative report based on the inversion results. This invention achieves a breakthrough in geological hydrogen migration from qualitative detection to quantitative tracking, significantly improving the reliability and accuracy of tracing, enabling real-time dynamic capture of the migration process, and providing precise decision-making basis for hydrogen exploration.

The invention relates to the technical field of nuclear facilities, and discloses a deuterium-tritium fusion device containing zoning method, device, equipment and medium, and the method comprises the steps: preliminarily dividing each tritium-related room into a plurality of tritium-related regions with different tritium containing grades based on a preset tritium-related concentration limit value; identifying a transition room between two tritium-related areas with adjacent spatial positions and adjacent tritium containing levels, and generating a plurality of different candidate zoning schemes by adjusting the tritium-related area to which the transition room belongs; for any candidate zoning scheme, determining system balance time of the candidate zoning scheme based on the space volume of each tritium-related region, the annual average tritium release rate and the preset net exhaust air rate; and identifying the shortest system balance time in each system balance time, and determining the candidate partitioning scheme corresponding to the shortest system balance time as a target partitioning scheme. The method has the beneficial effect that the dynamic response performance of the negative pressure ventilation system is optimized.

The invention relates to the technical field of isotopes, particularly provides an isotope battery coupled with C14 and a tritium radioactive source, and aims to solve the problems that in the prior art, a flexible film with radioactive elements cannot continue to use a conventional process, and introduction of the radioactive element tritium leads to degradation of a high polymer material and further leads to reduction of performance and service life. In order to achieve the purpose, the flexible isotope source of the isotope battery is prepared by the following steps: S1, obtaining 14C and 3H alkyne and a carbon source containing 14C; s2, enabling the 14C and 3H alkyne to react with a tritium source to obtain an intermediate product; and S3, adding three-dimensional graphene into the intermediate product, and carrying out polymerization reaction on the intermediate product. The flexible isotope source is prepared from the 14C and 3H alkyne, the 14C-containing carbon source, the tritium source and the three-dimensional graphene, so that the requirements of high tritium generation rate and flexibility of an isotope battery can be met, and the performance and the service life of a flexible radioisotope film are ensured.

A fusion system uses optical pulse shaping to control a fusion reaction by manipulating the quantum state of fluid fuel. The fusion system includes a reactor assembly that accepts fusion reaction fuel (e.g., hydrogen, hydrogen deuteride, deuterium, or deuterium-tritium, liquid water, liquid 17O-heavy water, etc?), and provides for conversion of high-energy fusion products to current. The fusion system may operate at ambient or low temperature (relative to ambient conditions), in the infrared band of the electromagnetic spectrum (e.g., between 700nm and 16000nm), and at moderate field intensities (e.g., between 109W / cm2 and 1017W / cm2), enabling efficient use of compact femtosecond laser sources and in-line, passive, or active optical elements for ultrafast pulse shaping. The net result is a field-robust electrical generator assembly that may be seamlessly integrated into existing electrical infrastructures.

In order to realize a fuel breeding unit that does not require the introduction of tritium recovery piping, a fuel breeding unit (1) converts kinetic energy of neutrons generated by nuclear fusion in plasma into thermal energy and generates tritium using neutrons. The fuel breeding unit (1) discharges the generated tritium directly into the reactor.

A breeder blanket is disposed within a stellarator. The breeder blanket defines a breeder zone through which a particle and a flow gas flows. The particle interacts with a neutron generated by a fusion reaction within the stellarator to generate heat. The flow gas absorbs the heat. A separator at an outlet of the breeder blanket separates the flow gas from the particle. The heat in the flow gas is used to generate power.