Helium-cooled Blanket Research Articles

Preliminary Assessment of Additively Manufactured Cooling Channel Performance for Helium-Cooled Blanket Concepts

Preliminary Assessment of Additively Manufactured Cooling Channel Performance for Helium-Cooled Blanket Concepts

Structure optimization of helium-cooled blanket for fusion reactor based on three-dimensional full-scale thermal-hydraulic analysis

Fusion energy is widely considered as the most promising method to solve the energy crisis, while the blanket is of vital importance in the fusion device due to its main function of thermal extraction and energy conversion. The high heat flux from plasma and the internal nuclear heat deposition are imposed on fusion blanket. Slice model or two-dimensional model is usually adopted for traditional design of fusion blanket, in which most side effects are neglected, introducing relatively large uncertainty for thermal removal capability of the blanket. In this study, the integral three-dimensional full-scale model is established based on the split-zone method for mesh generation. Based on the validated numerical method, the preliminary thermal-hydraulic analysis of the full-scale model shows that inadequate heat transfer and significant flow distribution difference are encountered for the initial blanket design, in which the full-scale effect is not considered. After that, the specific structure modifications are proposed for different blanket components to solve the abovementioned problems. The maximum temperature of first wall is reduced from 671°C to 639°C, which is under the material temperature limit of 650°C; the inverse V-shaped flow distribution in the cap is modified as more even flow in cap channels; the maximum temperature of the function zone in blanket is reduced from 999°C to 759°C, which is an important structure modification for the blanket. The integral three-dimensional full-scale thermal-hydraulic analysis is performed after the local structure modification. The results show that the function zone is cooled down obviously by the coolant in the side wall. Compared to the temperature limit of 900°C for lithium orthosilicate, there is a more rigorous limit for the beryllium material, since they share the common cooling capability layer by layer while the temperature limit for the latter is only 650°C. The three-dimensional full-scale thermal-hydraulic analysis in our research is valuable to evaluate the integral reliability of the fusion blanket.

3D Transient CFD Simulation of an In-Vessel Loss-of-Coolant Accident in the EU DEMO WCLL Breeding Blanket

The in-vessel Loss-of-Coolant Accident (LOCA) is one of the design basis accidents in the design of the EU DEMO tokamak fusion reactor. System-level codes are typically employed to analyse the evolution of these transients. However, being based on a lumped approach, they are unable to quantify localised quantities of interest, such as local pressure peaks on the vacuum vessel walls, to which the failure criteria are linked. To calculate local quantities, the 3D nature of the phenomenon needs to be considered. In this work, a 3D transient model of the in-vessel LOCA from a water-cooled blanket is developed. The model is implemented in the commercial CFD software STAR-CCM+. It simulates the propagation of the water jet in the vessel from the beginning of the accident, thus accounting for the phase change of the water, i.e., from the pressurised liquid phase to the vapour phase inside the vessel, being the latter at a much lower pressure than in the blanket coolant pipes. Due to the large pressure ratio (>1000), shocks are expected; therefore, an Adaptive Mesh Refinement (AMR) algorithm is employed. The physical models (in particular, the multiphase model) are benchmarked to a 2D reference problem before being applied to the 3D EU DEMO-relevant problem. The simulation results show that the pressure peaks in front of the vessel walls are not dangerous as they are below the design limit. The entire evolution of the water jet is followed up to the opening of the burst disks, in order to compare the average pressure evolution with that computed with system-level codes. A comparison with the in-vessel LOCA from a helium-cooled blanket is also carried out, showing that the accident evolution in the water case is less violent than in the helium case.

Toward a Fully Integrated Multiphysics Simulation Framework for Fusion Blanket Design

Fusion is an attractive clean-energy solution, thanks to its various advantages, such as reduced radioactivity, little high-level nuclear waste, ample fuel supplies, and increased safety. However, the harsh operating environment introduced by a complex fusion plasma system makes design and integration of fusion blankets incredibly challenging and time-consuming. This work focuses on developing a fully integrated multiphysics simulation framework based on an advanced open-source platform—the Multiphysics Object-Oriented Simulation Environment (MOOSE)—to alleviate the difficulties in fusion blanket design and integration. MOOSE is a massively parallel finite element/volume multiphysics simulation platform that has been widely adopted within the nuclear fission community. Even though fission and fusion are fundamentally different, they involve similar multiphysics phenomena. A fully integrated open-source multiphysics simulation framework tailored for the fusion blanket design will be implemented by leveraging the well-established multiphysics capabilities in MOOSE. Once successfully developed, this fully integrated framework will rapidly evaluate a blanket design concept and offer insights for subsequent iterations. As the first step, we will mainly aim to integrate neutronics analysis, system thermal hydraulics simulation, and full 3-D heat transfer calculations. The efficacy of the integrated framework will be verified using an innovative solid ceramic blanket design. While the project’s final goal is to enable a fully integrated multiphysics simulation platform for various fusion blanket concepts, this work, as a preliminary step, will mainly focus on a solid ceramic breeder helium-cooled blanket.

Highly underexpanded helium jet and its effect on dust resuspension under in-vessel loss of coolant accident of fusion reactor

During an in-vessel loss of coolant accident (LOCA) of a helium-cooled blanket in a fusion reactor, the high-pressure helium gas ingress into the vacuum vessel (VV) will produce highly underexpanded jet, resulting in the resuspension and migration of the radioactive dust deposited in the VV. In this study, the flow field in the VV and the influence of friction velocity were simulated and analyzed by using ANSYS Fluent code. In addition, the differences between in-vessel LOCA and loss of vacuum accident (LOVA) were compared. The results showed that the pressure of the VV reaches the pressure relief limit of the VV pressure suppression system rapidly after a LOCA. With the development of a LOCA, the Mach disk continued to move towards the equatorial break of the VV, its shape continued to shrink and the number of structures gradually increased. Compared with the LOVA, the LOCA had a higher friction velocity, resulting in a greater probability of dust resuspension and radioactive risk.

Titanium beryllide as an alternative to beryllium in nuclear and thermonuclear engineering, capabilities of UMP JSC in the technology development and beryllides products manufacture

For many decades, beryllium has been used as a structural element in nuclear installations as a moderator / breeder of fast neutrons. The consequence of neutron irradiation is a significant production of gas products in the form of helium and tritium, which leads to swelling and loss of strength properties of beryllium reflectors. The relatively low melting point of beryllium also imposes restrictions on the high-limit temperature regimes of the reactor core. As an alternative to pure beryllium, it is necessary to consider intermetallic compounds based on it, in particular titanium beryllide. Preliminary studies on the thermal desorption of helium and tritium from titanium beryllide have shown that this material has a much lower retention tendency and a lower release temperature. The higher melting point of titanium beryllide compared to pure beryllium is also an advantageous characteristic.Over the past years, UMP JSC, thanks to its research in this area, has achieved significant success in the development of technology for obtaining intermetallic billets and articles based on titanium and chromium beryllides. As a technology demonstrator, prototypes of structural elements of a helium-cooled blanket breeder module of the projected DEMO reactor were made by order of the Karlsruhe Institute of Technology, Germany.The advantages of titanium beryllide, as well as the success achieved in the production of billets and products from it, open up opportunities for a more extensive study of the nuclear, physical and mechanical properties of this material with the possibility of further use in nuclear technology, including thermonuclear reactors, and in high-temperature instrumentation.

Model Development and Transient Analysis of the HCPB BB BOP DEMO Configuration Using the Apros System Code

Extensive modeling and analytical work has been carried out considering the Helium-Cooled Pebble Bed Breeding Blanket (HCPB BB) Balance Of Plant (BOP) configuration of the Demonstration Power Plant (DEMO) using the Apros system code, developed by VTT Technical Research Centre of Finland Ltd. and Fortum. The integral plant model of the HCPB BB plant has been improved with respect to the blanket and steam generator models. Based on HCPB-BL2017 v1 data, reported in 2019, the blanket has been remodeled by separate Apros process components, dedicated to average inboard and outboard segments, where the power deposition scheme of the breeding units took into account the output of high-fidelity neutronic analyses. A new helical coil steam generator model has been developed for primary–secondary system coupling using CAD data provided by EUROfusion partner University of Palermo. Transient analyses have been performed with Apros on the plant configuration that utilizes a molten salt technology-based small Energy Storage System (ESS).

Optimization of the supercritical CO2 power conversion system based on the net efficiency under conditions of the pulse-operated fusion power reactor DEMO

Fusion power plants represent a new energy technology with specific features including multiple heat sources with different outlet temperature and power, fluctuating high self-consumption, and in the case of the demonstration fusion power plant, the pulse operation of the heat sources. The supercritical CO2 (S-CO2) Brayton cycle was applied and optimized under DEMO conditions with the nuclear net efficiency as the main optimization criterion. The optimization was performed for helium-cooled and water-cooled reactor blanket concepts, and for simple and recompression S-CO2 cycle layouts with four heat sources. The optimization process used a brute-force search problem-solving technique applied to multiparametric space. The model data was taken from the European model of the fusion power plant DEMO1 2019. When applying the simple S-CO2 cycle, the net efficiency of the model power plant with the helium-cooled and water-cooled blanket was found to be 19.5% and 12.3%, respectively, whereas when applying the recompression S-CO2 cycle, the net efficiency was found to be 16.4% and 8.1%, respectively. The simple S-CO2 cycle provided higher net efficiency for the model fusion power plants with the given configuration of multiple heat sources than the recompression S-CO2 cycle, and the model power plants using the helium-cooled blanket achieved higher net efficiency compared to the power plants using the water-cooled blanket despite significantly higher self-consumption. The application of the simple S-CO2 cycle allowed the model fusion power plant with the helium-cooled blanket to achieve the net efficiency higher than the reference net efficiency of 17.9% of the current DEMO model using the steam cycle. For the power plants with the water-cooled blankets, S-CO2 cycles were found to be less efficient than the steam cycles.

3D transient CFD simulation of an in-vessel loss-of-coolant accident in the EU DEMO fusion reactor

Accidental transients in fusion reactors, such as the in-vessel loss-of-coolant accident (LOCA) considered here, are generally analyzed using system-level codes. However, because of their lumped nature, such tools cannot predict local pressure and temperature values, which are instead of interest to assess the integrity of the vacuum vessel (VV) structures. It is then fundamental to prove that the system-level tools’ predictions are at least conservative. In this work, we analyze the helium flow inside the VV following a LOCA in the helium-cooled blanket of the EU DEMO, using a 3D transient computational fluid-dynamics (CFD) model implemented in the commercial STAR-CCM+ code. In view of the large pressure ratio, a hypersonic flow regime (Ma > 5) develops, with the formation of shock fronts requiring a high time and space resolution. The model is applied to compute the evolution of the pressure distribution inside the VV and then to compare it with the information provided by the system-level GETTHEM model. The predictions by the two models of the intervention time of the VV pressure suppression system are compared, showing that the 0D model is conservative.

Application of the best-estimate model calibration and prediction through experimental data assimilation methodology to the tests performed on a helium cooled First Wall Mock-up

Abstract Experiments are under progress at the Karlsruhe Institute of Technology (KIT) as part of the general strategy for the development of the DEMO Helium Cooled Pebble Bed (HCPB) design. These experiments are part of the program on the qualification of components and on the validation of system codes. In 2018, thermal-hydraulic investigations have been performed looking into the transient response of a helium-cooled blanket First Wall Mock-Up under Loss Of Flow Accident (LOFA) conditions. The results obtained through these investigations captured the quick temperature rise on the surfaces exposed to heat loads (such as those coming from the plasma), and provided a robust set of data for the validation of system codes such as RELAP5-3D. Then, the present work describes the validation activities performed on the RELAP5-3D code. This validation was performed employing also a Best-Estimate methodology called “Best-Estimate Model Calibration and Prediction through Experimental Data Assimilation methodology”. In a previous work, the same methodology was used to calibrate a RELAP5-3D model of the HELOKA-HP electrical heater under normal operation. In this regard, the present work marks the first application of this methodology to fast transients in the field of fusion reactors.

Progress in the design development of EU DEMO helium-cooled pebble bed primary heat transfer system

In the frame of the activities promoted and encouraged by the EURO-fusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently posed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, a very important role is played by the Breeding Blanket (BB) Primary Heat Transfer System (PHTS) as it is responsible to extract more than 80% of the fusion plasma power.In this framework, University of Palermo, Ansaldo Nucleare and CREATE have focused their work to improve thermal-hydraulic, safety and integration features of the Ex-Vessel PHTS for the Helium-Cooled Pebble Bed (HCPB) BB concept of DEMO.Starting from the outcomes obtained in 2016 that have allowed to highlight some criticalities which needed design changes, the paper describes progress and developments of the 2017 HCPB PHTS as well as the goals which have been achieved. The results of the research activity carried out show, in fact, i) an increase of the thermal-hydraulic performances, since a reduction in both total coolant inventory and total pressure drop has been respectively reached, ii) a potential improvement of the overall safety characteristics of the system. Nevertheless a critical assessment of these key parameters reveals that some issues are still open in terms of design integration and feasibility of the whole DEMO BoP for the helium-cooled blanket option, indicating that additional efforts are required to make this technology more attractive.

Design and Analysis of “Filling-Evacuating” High-Pressure Helium-Cooled Loop

The breeder blanket and divertor are crucial plasma facing components in a fusion reactor. The helium cooled blanket and divertor concepts have exhibited the best potential to come up to the highest safety requirements and therefore been chosen for the development object. As a result of high heat flux radiated from the plasma in the fusion reactor and high power density nuclear heat deposited by high-energy neutrons, the cooling of the first wall (FW) and the discharge of nuclear heat have become one of the major technical challenges. To demonstrate and verify the helium-cooled technology and tools of China test blanket module, and explore the feasibility and key technology of thermal hydraulics process of helium-cooled divertor, we have creatively adopted the “filling-evacuating” approach to design and fabricate the high-pressure helium-cooled loop (HPHCL), in which a mock-up of reduced-scale helium-cooled blanket module is designed and manufactured as a test section. Based on different experimental cases, the operating pressures of helium at mock-up range from 3 to 10 MPa and the maximum mass flow rate can reach up to 0.21 kg/s. In this paper, the design scheme of the HPHCL is presented, and the key issues of engineering manufacture and the test cases are calculated and analyzed. The helium gas flow and the heat transfer are calculated according to the test working conditions of the referenced International Thermonuclear Experimental Reactor TBM’s FW surface heat flux, using ANSYS fluid dynamics software FLUENT. The results will provide support for the follow-up fabrication of test system and implementation of the tests.

Preliminary evaluation of the expansion system size for a pressurized gas loop: application to a fusion reactor based on a helium-cooled blanket

Preliminary evaluation of the expansion system size for a pressurized gas loop: application to a fusion reactor based on a helium-cooled blanket

Thermal-Hydraulic Analysis for Conceptual Design of Korean HCCR TBM Set

With a conceptual design of Korean helium-cooled ceramic reflector test blanket module (TBM), including TBM shield for testing in ITER, thermal-hydraulic analyses are performed with a conventional CFD code, ANSYS-CFX v14.5, for the electromagnetic module and the integral module, including the TBM shield. With the same model and meshes, according to the ITER operation conditions of the H/He and D-T phases, the temperature distribution, flow rate, and pressure drop are investigated to meet the design requirements, and the temperature data are directly provided to a mechanical analysis.

Conceptual Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR

The China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak reactor proposed by the China National Integration Design Group. The aim is to develop China’s next-step fusion device. A helium-cooled ceramic breeder (HCCB) blanket concept has been put forward by the blanket integration design team of the Institute of Plasma Physics of the Chinese Academy of Sciences. The present blanket configuration is based on the upper port dimension and maintenance scheme. The HCCB blanket comprises a U-shaped first wall, cap, breeder unit, middle plate, and back plate. The breeder unit includes a cooling plate, beryllium pebble bed, and lithium silicate pebble bed. The structure of the HCCB blanket concept is obviously different from the ITER HCCB test blanket module of China. The CFETR HCCB blanket will meet the requirements of CFETR long-pulse or steady-state operation with a duty cycle time greater than 0.3. The maintenance scheme of the blanket is introduced in this paper. Three-dimensional neutronic analysis results show that the tritium breeding ratio can satisfy the design requirement. Besides, the thermal-hydraulic behavior of the first wall has been studied using ANSYS CFX code.

Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept

As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters.

Tritium Transport Issues for Helium-Cooled Breeding Blankets

Tritium mobility through breeding blanket (BB) and steam generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason, it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modeling and parametric issues. In this paper, the main permeation issues are identified and possible solutions are discussed to address the modeling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled BBs: 1) helium-cooled pebble beds and 2) helium-cooled lithium-lead. For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC to define proper and feasible tritium mitigation techniques, which are needed to keep the tritium losses lower than the allowable environmental release (i.e., 20 Ci/d).

Static Neutronics Analyses of Hellium-Cooled Solid Breeder Blanket

A Monte Carlo N-particle transport code was used to study physics of the helium-cooled solid-breeder tritium breeding blanket in the Chinese Fusion Engineering Thermal Reactor （CFETR）for various volume ratio of the neutron multiplier and tritium breeder and various thickness of the first wall. A sandwich-type of Be and loading model is used to analyze the compact of volume ratio and the thickness of the first wall for the tritium breeding rate. The results of different volume ratio models show that the tritium breeding ratio would reach 1.51 for volume ratio from 2 to 5.And the results of the different first wall thickness show that the upper limit of the thickness should be 33mm to keep the tritium self-sustain.

Requirements for helium cooled pebble bed blanket and R&D activities

This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R&D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM.The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R&D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R&D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.

Fabrication of Li4SiO4 Pebbles Using Slurry Droplet Wetting Method for Solid Breeding Material

Lithium orthosilicate (Li4SiO4) pebbles are the primary candidate breeder for the Korean Helium-Cooled Solid Breeder test blanket module because of their superior overall performance qualities such as reasonable lithium atom density, favorable tritium release behavior, low activation property, and chemical stability. The aim of this study is to develop a fabrication method for Li4SiO4 pebbles with high sphericity and stable micro-crystalline structure. Li4SiO4 slurry was prepared by mixing Li4SiO4 powder and polyvinyl alcohol (PVA) solution. The Li4SiO4 gel-spheres were made by dropping slurry mixture into hydrogen peroxide (H2O2) solution. The gel-spheres were floated up to the surface of the H2O2 solution after staying at the bottom for a while and then aged on the surface of the H2O2 solution. The gel-spheres were dried at room temperature and sintered at several temperatures for 4 hours in air atmosphere. The sintered pebbles kept the shape of spheres during the sintering process in spite of shrinkage. Furthermore, the pebbles had relatively high sphericity. Also, the effects of PVA content and sintering temperature on the properties of the Li4SiO4 pebbles were investigated.