Search Results (21)

In a tritium-breeding blanket of a fusion reaction, helium, used as a tritium-purging gas, will purge the tritium breeder pebble beds to extract the tritium in blanket. The purge gas flow characteristics will affect the tritium extraction efficiency. The effect of the fixed wall on the pebble packing structures and purge gas flow characteristics was investigated by combining the discrete element method (DEM) and computational fluid dynamics (CFD) method. The results indicate that the fixed wall leads to a regular packing of the pebbles adjacent to the fixed wall in association with drastic fluctuations in the porosity of the pebble bed, which can affect the purge gas flow behaviors. Further analyses of helium flow behaviors show that the helium pressure in the pebble bed decreases in a linear manner along the flow direction, whereas the pressure drop gradient of helium increases gradually with an increase in the packing factor. The reduction in porosity in the pebble bed leads to a notable escalation in helium flow velocity. Concerning the direction perpendicular to the helium gas flow, the evolution of the cut-plane averaged velocity of helium is similar to that of the porosity, except in the region immediately adjacent to the wall. The pressure drop and flow characteristics obtained in this study can serve as input for the thermohydraulic analysis of the tritium blowing systems in the tritium-breeding blanket of a fusion reactor. Full article

The flow characteristics of purge gas helium in the pebble bed of the tritium breeding blanket are important in analyzing the tritium purging process and optimizing the design of the solid breeder blanket. Therefore, the flow characteristics of helium gas in randomly packed pebble beds are investigated experimentally with a focus on the analysis of the pressure loss distribution. The results show that gas velocity, bed dimension, and pebble diameters have an obvious influence on the helium flow characteristics in pebble beds. With the increase in the inlet helium gas velocity, the pressure drop gradient of helium in the pebble bed gradually increased. With increases in the pebble bed dimension, the pressure drop gradient of helium in the pebble bed gradually increased. With the increase in the pebble diameter, the pressure drop gradient gradually decreased. In addition, the effect of temperature on the pressure drop of helium in the pebble bed was also preliminarily investigated. The pressure drop gradient of the helium through the pebble bed obviously increased with the increase in the helium and the bed temperature. The experimentally obtained pressure loss characteristics can be used for the validation of the simulation of a blanket pebble bed and as input parameters in the thermo-hydraulic analysis of solid-tritium breeder blankets. Full article

Research and development of the DEMOnstration power plant (DEMO) breeder blanket (BB) has been performed in recent years based on a predefined DEMO tritium breeding ratio (TBR) requirement, which determines a loss of wall surface due to non-breeding in-vessel components (IVCs) which consume plasma-facing wall surface and do not contribute to the breeding of tritium. The integration of different IVCs, such as plasma limiters, neutral beam injectors, electron cyclotron launchers and diagnostic systems, requires cut-outs in the BB, resulting in a loss of the breeder blanket volume, TBR and power generation, respectively. The neutronic analyses presented here have the goal of providing an assessment of the TBR losses associated with each IVC. Previously performed studies on this topic were carried out with simplified, homogenized BB geometry models. To address the effect of the detailed heterogeneous structure of the BBs on the TBR losses due to the inclusion of the IVCs in the tokamak, a series of blanket geometry models were developed for integration in the latest DEMO base model. The assessment was performed for both types of BBs currently developed within the EUROfusion project, the helium-cooled pebble bed (HCPB) and water-cooled lead–lithium (WCLL) concepts, and for the water-cooled lead and ceramic breeder (WLCB) hybrid BB concept. The neutronic simulations were performed using the MCNP6.2 Monte Carlo code with the Joint Evaluated Fission and Fusion File (JEFF) 3.3 data library. For each BB concept, a 22.5° toroidal sector of the DEMO tokamak was developed to assess the TBR and nuclear power generation in the breeder blankets. For the geometry models with the breeder blanket space filled only with blankets without considering IVCs, the results of the TBR calculations were 1.173, 1.150 and 1.140 for the HCPB, WCLL and WLCB BB concepts, respectively. The TBR impact of all IVCs and the losses of the power generation were estimated as a superposition of the individual effects. Full article

Two-phase lithium ceramic Li₂TiO₃-Li₄SiO₄ is considered as a tritium multiplier for use in the solid blanket of fusion reactors. To date, the most accurate understanding of the processes of tritium and helium production and release occurring in the breeder blanket materials under neutron irradiation can only be obtained from experiments in fission research reactors. At that, irradiations in vacuum give the possibility to register even very fast gas release processes (bursts) from the ceramics’ voids and pores, although it reduces the thermal conductivity of the pebble bed. The purpose of this work was to simulate the heating of mono-sized pebble bed (1 mm in diameter) of two-phase lithium ceramic 25 mol%Li₂TiO₃+75 mol%Li₄SiO₄ in an ampoule device during neutron irradiation at the WWR-K research reactor under vacuum conditions, and to determine experimental parameters in order to prevent heating of the lithium ceramics up to the Li₄SiO₄-Li₂SiO₃ phase transition temperatures (>900 °C). For the first time, it was obtained that the effective thermal conductivity of a 1 mm mono-sized pebble bed of 25 mol%Li₂TiO₃+75 mol%Li₄SiO₄ significantly decreases (four times) when it is irradiated with neutrons in a vacuum (at a helium pressure of approximately 10 Pa), compared to a similar calculation at 100 kPa of helium (when the He sweep is used). It was concluded that it is difficult to evaluate the maximal temperature of the ceramics in the capsule by measuring the temperature of its outer metal wall (according to thermocouple readings) without using the results of thermophysical calculations for each type of ceramic, taking into account its quantity, specific heat release and pebble size(s). To control the temperature of the ceramics during an irradiation experiment in a vacuum, an in-capsule thermocouple should be used, placed in the center of the pebble bed. Measuring the temperature of the pebble bed based on the capsule wall temperature can lead to overheating of the ceramics and phase changes. Full article

The operation of fusion reactors is based on the reaction that occurs when two heavy hydrogen isotopes, deuterium and tritium, combine to form helium and a neutron with an energy of 14.1 MeV D + T → He + n. For this reaction to occur, it is necessary to produce tritium in the facility itself, as tritium is not common in nature. The generation of tritium in the facility is a key function of the breeder blanket. During the operation of a D–T fusion reactor, high-energy tritium is generated as a result of the ⁶Li(n,α)T reaction in a lithium-containing ceramic material in the breeder blanket. Lithium metatitanate Li₂TiO₃ is proposed as one of the promising materials for use in the solid breeder blanket of the DEMO reactor. Several concepts for test blanket modules based on lithium ceramics are being developed for testing at the ITER reactor. Lithium metatitanate Li₂TiO₃ has good tritium release parameters, as well as good thermal and thermomechanical characteristics. The most important property of lithium ceramics Li₂TiO₃ is its ability to withstand exposure to long-term high-energy radiation at high temperatures and across large temperature gradients. Its inherent thermal stability and chemical inertness are significant advantages in terms of safety concerns. This study was a continuation of research regarding tritium and helium release from lithium metatitanate Li₂TiO₃ with 96% ⁶Li during irradiation at the WWR-K research reactor using the vacuum extraction method. As a result of the analysis of experiments regarding the irradiation of lithium metatitanate in vacuum conditions, it has been established that, during irradiation, peak releases of helium from closed pores of the ceramics are observed, which open during the first 7 days of irradiation. The authors assumed that the reasons samples crack are temperature gradients over the ceramic sample, resulting from the internal heating of pebbles under the conditions of their vacuum evacuation, and contact with the bottom of the evacuated capsule. The temperature dependence of the effective diffusion coefficient of tritium in ceramics at the end of irradiation and the parameters of helium effusion were also determined. Full article

An emerging breeding blanket that fulfills performance criteria, meets the safety requirements, and is reliable enough to meet the plant availability is a challenging issue that assumes complex studies involving numerous neutronic analyses based on the Monte Carlo simulations with MCNP code. Two different concepts are now candidates to be implemented as a driver blanket for DEMO fusion reactor: WCLL (Water-Cooled Lithium Lead) and HCPB (Helium-Cooled Pebble Bed). The current R&D work within the EUROfusion DEMO project is concentrated on a search for the new water-cooled blanket layouts: a deep upgrade of the WCLL blanket to ensure a sufficient tritium breeding capability and an elaboration of the hybrid concept coupling technological advantages of water coolant, lead neutron multiplier, and ceramic breeder. To this end, very detailed, fully heterogeneous MCNP geometry models were developed for the newest designs of the WCLL-db (WCLL-double bundle) and WLCB (Water-cooled liquid Lead Ceramic Breeder) DEMO blankets to verify the new engineering solutions. This makes rigorous calculations possible to find an optimal breeder blanket layout. The basic response, tritium breeding ratio (TBR), was assessed for both concepts, and it appeared to be TBR = 1.16 for the WCLL-db and TBR ≤ 1.13 for the WLCB DEMOs, respectively. Several geometry layouts of the WLCB breeder blanket were investigated to reach the TBR sufficient for a sustainable tritium fuel cycle. Two promising novel solutions were suggested to enhance the tritium breeding performance of the WLCB blanket and to achieve TBR ≥ 1.16: heavy water coolant and an advanced breeder ceramic. Various nuclear safety aspects of the technologies utilized in both blanket concepts are addressed in this work to facilitate engineering decisions aimed at the consolidated blanket design for the upcoming DEMO reactor. Full article

The Helium Cooled Pebble Bed (HCPB) breeding blanket is one of the two driver-blanket candidates for the European fusion demonstration power plant (EU DEMO) within the framework of the EUROfusion Consortium. As the EU DEMO program is going, testing of mockups becomes increasingly important. In this article, the engineering design of a first-ever breeder zone mockup of the EU DEMO HCPB breeding blanket is reported. The mockup will be tested in the high-pressure, high temperature, helium facility (HELOKA) at Karlsruhe Institute of Technology. This mockup will act as a test rig to validate heat transfer correlations, CFD software, and thermal hydraulics systems codes. As pressure equipment, the mockup shall conform to the latest European Union Pressure Equipment Directive 2014/68/EU. The design description, rationale and test matrix, and corresponding analyses are discussed and presented. Full article

The fusion program in the Karlsruhe Institute of Technology (KIT) leads the R&D of the DEMO helium-cooled pebble bed (HCPB) breeding blanket within the work package breeding blanket (WPBB) of the Eurofusion Consortium in the European Union (EU). A new design of the HCPB breeder zone, with a layout inspired by a nuclear reactor fuel rod arrangement, was developed recently and called the fuel-breeder pin concept. In addition, a mock-up (MU) of this fuel-breeder pin was designed and manufactured at KIT in order to test and validate its thermal–hydraulic performance. This paper reports on the results of the first experimental campaign dedicated to the fuel-breeder pin MU testing that was performed in the Helium Loop Karlsruhe (HELOKA) facility. The paper presents: (i) the integration of the fuel-breeder pin MU into the HELOKA loop including considerations of the experimental set-up, (ii) an overview of the plan for the experimental campaigns, and (iii) a discussion of the experimental results with a focus on aspects relevant for the validation of the thermal–hydraulic design of the HCPB breeder zone. Full article

The main components of a liquid breeder blanket in a fusion power reactor are lead lithium alloy (PbLi) and the steel structure in which the liquid is enclosed (EUROFER). Several compatibility tests have shown that structural materials always suffer from corrosion attacks. The governing mechanism can be attributed to the dissolution of the steel by the liquid breeder and is strongly related to the PbLi chemistry, velocity profile, and temperature. A new facility, CiCLo-C (CIEMAT Corrosion Loop, Internally Coated), is dedicated to the study of corrosion in materials under the severe breeding blanket condition. An effort was made to design an experimental facility with a specific test section able to work at quite ambitious operation parameters: up to 550 °C and a 1 m/s flow of PbLi. Furthermore, an innovative tantalum coating was introduced in the whole loop to avoid impurities coming from the pipeline, which can disturb the measurements, and to better preserve the installation. Full article

Tritium permeation into the Primary Heat Transfer System (PHTS) of DEMO and ITER reactors is one of the challenging issues to be solved in order to demonstrate the feasibility of nuclear fusion power plants construction. Several technologies were investigated as antipermeation and corrosion barriers to reduce the tritium permeation flux from the breeder into the PHTS. Within this frame, alumina coating manufactured by Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) are two of the main candidates for the Water Cooled Lithium Lead (WCLL) Breeder Blanket (BB). In order to validate the performance of the coatings on relevant WCLL BB geometries, a mock-up was designed and will be characterized in an experimental facility operating with flowing lithium-lead, called TRIEX-II. The present work aims to illustrate the preliminary engineering design of a WCLL BB mock-up in order to deeply investigate permeation of hydrogen isotopes through PHTS water pipes. The permeation tests are planned in the temperature range between 330 and 500 °C, with hydrogen and deuterium partial pressure in the range of 1–1000 Pa. The hydrogen isotopes transport analysis carried out for the design and integration of the mock-up in TRIEX-II facility is also shown. Full article

The article is dedicated to the study of the phase formation processes in Li₂ZrO₃ ceramics obtained by the method of solid phase synthesis. Interest in these types of ceramics is due to their great potential for use as blanket materials in thermonuclear reactors, as well as being one of the candidates for tritium breeder materials. Analysis of the morphological features of the synthesized ceramics depending on the annealing temperature showed that the average grain size is 90–110 nm; meanwhile the degree of homogeneity is more than 90% according to electronic image data processing results. The temperature dependences of changes in the structural and conducting characteristics, as well as the phase transformation dynamics, have been established. It has been determined that a change in the phase composition by displacing the impurity LiO and ZrO₂ phases results in the compaction of ceramics, as well as a decrease in their porosity. These structural changes are due to the displacement of LiO and ZrO₂ impurity phases from the ceramic structure and their transformation into the Li₂ZrO₃ phase. During research, the following phase transformations that directly depend on the annealing temperature were established: LiO/ZrO₂/Li₂ZrO₃ → LiO/Li₂ZrO₃ → Li₂ZrO₃. During analysis of the obtained current-voltage characteristics, depending on the annealing temperature, it was discovered that the formation of the Li₂ZrO₃ ordered phase in the structure results in a rise in resistance by three orders of magnitude, which indicates the dielectric nature of the ceramics. Full article

According to the most recently revised European design strategy for DEMO breeding blankets, mature concepts have been identified that require a reduced technological extrapolation towards DEMO and will be tested in ITER. In order to optimize and finalize the design of test blanket modules, a number of issues have to be better understood that are related to the magnetohydrodynamic (MHD) interactions of the liquid breeder with the strong magnetic field that confines the fusion plasma. The aim of the present paper is to describe the state of the art of the study of MHD effects coupled with other physical phenomena, such as tritium transport, corrosion and heat transfer. Both numerical and experimental approaches are discussed, as well as future requirements to achieve a reliable prediction of these processes in liquid metal blankets. Full article

A new type of Water Lead Lithium Cooled (WLLC) blanket that adopts the modular design scheme, water cooling the structure components, liquid PbLi as breeder and coolant, and SiC as the thermal insulator between PbLi and structures is under development as a candidate blanket concept for the Chinese Fusion Engineering Test Reactor (CFETR). Based on a poloidal-radial slice model, thermal hydraulic analysis is performed for this blanket to validate the feasibility of design goals. Results show that the present design can achieve the outlet temperature in the range of 600–700 °C, with all the material temperatures safely below the upper limits. A series of sensitivity analyses are also carried out. It indicates that the thermal conductivity (TC) of SiC would have a significant influence on the temperature field, streamlines and pressure drop; that is, lower TC of SiC can maintain the temperature of PbLi at a high level, and induce an increased number of vortices in the liquid PbLi flow as well as a larger pressure drop. On this basis, the joint effects of the TC of SiC and inlet velocity on the performance of blanket thermal hydraulics are analyzed, then the so-called “attainable region” is proposed. Finally, optimization design studies are carried out by decreasing the width of the front channel. Comparison results show that the present design is the most reasonable. Full article

Magnetohydrodynamic (MHD) phenomena, due to the interaction between a magnetic field and a moving electro-conductive fluid, are crucial for the design of magnetic-confinement fusion reactors and, specifically, for the design of the breeding blanket concepts that adopt liquid metals (LMs) as working fluids. Computational tools are employed to lead fusion-relevant physical analysis, but a dedicated MHD code able to simulate all the phenomena involved in a blanket is still not available and there is a dearth of systems code featuring MHD modelling capabilities. In this paper, models to predict both 2D and 3D MHD pressure drop, derived by experimental and numerical works, have been implemented in the thermal-hydraulic system code RELAP5/MOD3.3 (RELAP5). The verification and validation procedure of the MHD module involves the comparison of the results obtained by the code with those of direct numerical simulation tools and data obtained by experimental works. As relevant examples, RELAP5 is used to recreate the results obtained by the analysis of two test blanket modules: Lithium Lead Ceramic Breeder and Helium-Cooled Lithium Lead. The novel MHD subroutines are proven reliable in the prediction of the pressure drop for both simple and complex geometries related to LM circuits at high magnetic field intensity (error range $\pm 10$%). Full article

A tritium breeding blanket (TBB) is an essential component in a fusion reactor, which has functions of tritium breeding, energy generation and neutron shielding. Tritium breeding ratio (TBR) is a key parameter to evaluate whether the TBB could produce enough tritium to achieve tritium self-sufficiency (TBR > 1) for a fusion reactor. Current codes or software struggle to meet the requirements of high efficiency and high automation for neutronic optimization of the TBB. In this paper, the multiphysics coupling and automatic neutronic optimization method study for a solid breeder TBB is performed, and a corresponding code is developed. A typical module of China fusion engineering test reactor (CFETR) helium cooled ceramic breeder (HCCB) TBB was selected, and a 3D neutronics model of an initial scheme is developed. The automatic neutronic optimization was performed by using the developed code for verification. Results indicate that the TBR could increase from 1.219 to 1.282 (~5.17% improvement), and that the maximum temperature of each type of material in the optimized scheme is below the allowable temperature. It is of great scientific significance and engineering value to explore and study the algorithm for automatic neutronic optimization and the code development of the TBB. Full article